ORAU TEAM

Dose Reconstruction

Project for NIOSH

Oak Ridge Associated Universities I Dade Moeller I MJW Technical Services

Page 1 of 150

DOE Review Release 10/26/2009

Document Title:

Los Alamos National Laboratory – Occupational 

Internal Dose

Document Number: ORAUT-TKBS-0010-5 

Revision: 01 

Effective Date: 10/15/2009

Type of Document TBD

Supersedes: Revision 00 

Subject Expert(s): Jack E. Buddenbaum, Donald N. Stewart, Rowena S. Argall

Site Expert(s): N/A

Approval: Signature on File Approval Date: 09/26/2009

Donald N. Stewart, Document Owner

Concurrence: Signature on File Concurrence Date: 09/27/2009

John M. Byrne, Objective 1 Manager

Concurrence: Keith A. McCartney Signature on File for Concurrence Date: 09/28/2009

Edward F. Maher, Objective 3 Manager

Concurrence: Signature on File Concurrence Date: 09/28/2009

Kate Kimpan, Project Director

Approval: Signature on File Approval Date: 10/15/2009

James W. Neton, Associate Director for Science

 New Total Rewrite Revision Page Change

FOR DOCUMENTS MARKED AS A TOTAL REWRITE, REVISION, OR PAGE CHANGE, REPLACE THE PRIOR 

REVISION AND DISCARD / DESTROY ALL COPIES OF THE PRIOR REVISION.
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 2 of 150

PUBLICATION RECORD

EFFECTIVE 

DATE

REVISION 

NUMBER DESCRIPTION

12/21/2004 00 New Technical Basis Document for the Los Alamos National 

Laboratory – Occupational Internal Dose. Incorporates formal 

internal and NIOSH review comments. Initiated by Jack E. 

Buddenbaum.

10/15/2009 01 Incorporates information from documents not available at the time 

the original document was published. The current information 

available for uranium has been incorporated into Section 5.2.4. 

Clarification of MDAs, UNAA/DNAA analysis results, guidance for 

selection of enrichment, and a description of the new LANL Bioassay 

Repository database content have been included. Appropriate 

references have been added. For claims involving uranium UNAA 

bioassay results: Because of the ambiguity in the interpretation of 

the uranium bioassay results when results were transferred into the 

Bioassay Repository, the guidance in the TBD might not have been 

interpreted consistently by DRs and PRs. UNAA/DNAA data 

provided by LANL is for DU and EU rather than U-238 and U-235. A 

table providing an aged “pure” 238Pu mixture has been provided for 

use when exposures to heat source technology plutonium is 

suspected. Changes to the mixture might affect the total dose for 

plutonium intakes evaluated as ”pure” 238Pu. In addition, the Tiger 

Team Assessment (DOE 1991), the Human Experiments and the 

details of the RaLa SEC Report (NIOSH 2006) have been included in 

the revision. Incorporates formal internal and NIOSH review 

comments. Adds Attributions and Annotations section. A description 

of the class of employees added to the SEC for SEC Petition-00051 

was added to the text describing the class of employees added under 

SEC Petition-00061 (employees associated with RaLa operations). 

A new Section 5.2.4.6 was added describing the limitations on 

internal dose reconstruction for employees in the non-excepted TAs 

(as listed in the petition). Added statement about only partial dose 

reconstructions are possible for most TAs. A new bullet was added 

in Section 5.7.2, General Guidance, indicating that only partial 

internal dose reconstructions are possible. Incorporates methods for 

estimating maximum doses in accordance with the evaluation report 

for Petition SEC-00109, changes to address comments by OCAS 

commenters, updated dose reconstruction methodologies and 

references, and addition of an algorithm to guide dose reconstructors 

for dose calculation and dose assignment. Constitutes a total rewrite 

of the document. Training required: As determined by the Objective 

Manager. Initiated by Donald N. Stewart.

Trademark Information Microsoft and Excel are registered trademarks of Microsoft Corporation in 

the United States and/or other countries.
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 3 of 150

TABLE OF CONTENTS

SECTION TITLE

Acronyms and Abbreviations ...............................................................................................................

PAGE

8

5.1 Introduction ............................................................................................................................11

5.1.1 Purpose......................................................................................................................12

5.1.2 Scope.........................................................................................................................12

5.1.3 Classes of Employees Added to the Special Exposure Cohort ................................... 13

5.1.4 Background of Internal Exposures..............................................................................13

5.1.4.1 Early Research with Humans....................................................................... 14

5.1.4.2 Early Safety Efforts ...................................................................................... 14

5.1.4.3 Bioassay Program........................................................................................ 16

5.1.4.3.1 Individual Bioassay Results ........................................................ 17

5.1.4.3.2 UPPU Club ................................................................................. 18

5.1.5 Unmonitored Workers and Unreported Bioassay Results ........................................... 18

5.1.5.1 Workers Before 1949 ................................................................................... 18

5.1.5.2 Short-Term Workers..................................................................................... 19

5.1.5.3 Zia Company and Other Contract Service Workers...................................... 19

5.1.5.4 RaLa Project Operations, 1944 to 1963 ....................................................... 20

5.1.6 Overview of Internal Dose Reconstruction Process for Los Alamos National 

Laboratory ..................................................................................................................20

5.2 In Vitro Bioassay Minimum Detectable Amounts, Counting Methods, and Reporting 

Protocols for Primary Radionuclides ...................................................................................... 23

5.2.1 Plutonium ...................................................................................................................23

5.2.1.1 Respiratory Tract Absorption Type............................................................... 24

5.2.1.2 Sample Collection Procedures ..................................................................... 24

5.2.1.3 Missed Intakes............................................................................................. 25

5.2.1.4 Routine Sample Frequencies ....................................................................... 26

5.2.1.5 Sample Analysis Procedures ....................................................................... 26

5.2.1.6 Plutonium Isotopic Ratios (Mixtures)............................................................ 29

5.2.1.7 Correcting for Urinalysis Volume.................................................................. 32

5.2.1.8 Calculating Picocuries per 24 Hours for Urine Bioassay Results.................. 32

5.2.1.9 Detection Sensitivities, Reporting Limits, and Tolerance Limits.................... 32

5.2.1.10 Validation of Samples .................................................................................. 33

5.2.1.11 Sampling Protocol for Special Accidental Exposure ..................................... 34

5.2.1.12 Nasal Swipes ............................................................................................... 34

5.2.1.13 Computer-Based Calculations and Sorting................................................... 34

5.2.1.14 LANL Bioassay Data Repository .................................................................. 35

5.2.2 Americium ..................................................................................................................35

5.2.2.1 Minimum Detectable Activities ..................................................................... 36

5.2.3 Tritium ........................................................................................................................36

5.2.3.1 Organically Bound Tritium............................................................................ 37

5.2.3.2 Metal Tritides ............................................................................................... 37

5.2.3.3 Analytical Techniques .................................................................................. 37

5.2.3.4 Routine Sampling......................................................................................... 38

5.2.4 Uranium......................................................................................................................38

5.2.4.1 Sampling Protocol........................................................................................ 40

5.2.4.2 Uranium Analysis Techniques...................................................................... 40

5.2.4.3 Delayed Neutron Counting........................................................................... 41
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 4 of 150

5.2.4.4 Radiometric Alpha Spectroscopy.................................................................. 43

5.2.4.5 Potential Missed Intakes .............................................................................. 43

5.2.4.6 Excretion of Environmental Levels of Uranium............................................. 43

5.2.5 Polonium ....................................................................................................................44

5.2.6 Other Limited-Exposure Radionuclides.......................................................................45

5.3 In Vivo bioassay Minimum Detectable Activities, Analytical Methods, and Reporting 

Protocols................................................................................................................................45

5.3.1 Whole-Body Counters ................................................................................................46

5.3.1.1 Minimum Detectable Activities and Decision Levels..................................... 46

5.3.1.2 Cesium-137 and Other Intakes from Fallout................................................. 49

5.3.2 Lung Burdens .............................................................................................................50

5.3.3 Wound Monitoring ......................................................................................................52

5.4 Interferences and Uncertainties.............................................................................................. 52

5.4.1 Discrepancies in Results ............................................................................................52

5.4.2 Contamination of Samples..........................................................................................53

5.4.3 Uncertainties ..............................................................................................................53

5.4.3.1 In Vivo Counting........................................................................................... 53

5.4.3.2 In Vitro Measurements................................................................................. 53

5.5 Unmonitored Intakes..............................................................................................................54

5.5.1.1 Energy Employee Interview.......................................................................... 54

5.5.1.2 Survivor or Less-Detailed Interview.............................................................. 54

5.5.1.3 Energy Employee, Science Job Category Interview ..................................... 54

5.5.1.4 Initial Case File ............................................................................................ 54

5.5.2 Unmonitored Intakes of Primary Radionuclides ..........................................................54

5.5.2.1 Plutonium..................................................................................................... 55

5.5.2.2 Polonium...................................................................................................... 55

5.5.2.3 Uranium ....................................................................................................... 55

5.5.2.4 Tritium.......................................................................................................... 56

5.5.3 Unmonitored Intakes of LANL Radionuclides other than Primary Radionuclides ........ 56

5.5.3.1 Actinium-227................................................................................................ 57

5.5.3.2 Americium.................................................................................................... 58

5.5.3.3 Curium-244.................................................................................................. 58

5.5.3.4 Neptunium-237 ............................................................................................ 59

5.5.3.5 Protactinium-231.......................................................................................... 59

5.5.3.6 Thorium........................................................................................................ 60

5.5.3.7 Thorium-230 ................................................................................................ 60

5.5.3.8 Thorium-232 ................................................................................................ 60

5.5.3.9 Mixed Fission and Activation Products......................................................... 61

5.6 Attributions and Annotations .................................................................................................. 61

References ........................................................................................................................................63

Glossary ............................................................................................................................................74
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 5 of 150

ATTACHMENT A, OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS ................. 78

A.1 Scope.....................................................................................................................................78

A.2 Codes Used in Bioassay and Internal Dose Records ............................................................. 78

A.2.1 Los Alamos Bioassay Data Repository Codes............................................................78

A.3 Locations and Typical Radionuclides ..................................................................................... 80

A.4 Computer Code for Validating Plutonium Bioassay Samples ................................................. 80

A.4.1 Validations Techniques PUQFUA2.............................................................................80

A.4.2 PUQFUA Primary Validation Technique .....................................................................92

A.5 Summary of Computer Codes................................................................................................ 93

A.6 Current and Historical In Vitro Bioassay Programs................................................................. 95

A.7 Estimation of Date of Intake ................................................................................................... 95

A.8 Routine Sampling Procedures................................................................................................ 96

A.9 Summary of Analytical Techniques for Plutonium .................................................................. 96

A.10 Example Bioassay Results..................................................................................................... 96

A.11 Routine Sampling Frequency ................................................................................................. 97

A.12 Sampling Protocol for Accidental Exposures.......................................................................... 99

A.13 Summary of In Vitro and In Vivo Sensitivities ......................................................................... 99

A.14 In Vivo Bioassay Routine Frequency.................................................................................... 103

A.15 Human Experimentation....................................................................................................... 103

ATTACHMENT B, AIR CONCENTRATIONS IN SELECTED BUILDINGS ...................................... 114

ATTACHMENT C, BIOASSAY FOR RADIONUCLIDES OTHER THAN 

PRIMARY RADIONUCLIDES ............................................................................ 119

ATTACHMENT D, RESPIRATORY PROTECTION PROGRAM...................................................... 124

ATTACHMENT E, PROGRAMMATIC ISSUES TO BE REVIEWED FOR 

EVALUATING FOR POTENTIAL UNMONITORED INTAKES ........................... 126

ATTACHMENT F, DESCRIPTIONS OF SOME INCIDENTS RESULTING IN 

INTERNAL DOSES............................................................................................145

ATTACHMENT G, SCALING OF COWORKER DOSE INTAKES CONSISTENT 

WITH TYPE AND DURATION OF EXPOSURES .............................................. 149
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 6 of 150

LIST OF TABLES

TABLE TITLE

5-1 Historical in vitro bioassay, primary radionuclides................................................................. 

PAGE

23

5-2 Historical in vitro bioassay not available in claim records...................................................... 23

5-3 Plutonium respiratory tract absorption type........................................................................... 24

5-4 Activity composition of nominal 3% plutonium mixture .......................................................... 27

5-5 Activity composition of reference weapons-grade 6% plutonium mixture .............................. 27

5-6 Activity composition of reference fuel-grade (12%) plutonium mixture .................................. 28

5-7 Plutonium bioassay sensitivity as listed in procedures and reports ....................................... 30

5-8 “Pure” 238Pu mixture.............................................................................................................. 31

5-9 Fresh plutonium mixture ....................................................................................................... 31

5-10 Americium-241 bioassay techniques and sensitivities........................................................... 36

5-11 Tritium urine bioassay sensitivity levels ................................................................................ 38

5-12 Nominal weight composition and fractional activities for D-38 uranium ................................. 39

5-13 Nominal weight composition and fractional activities for OR-93 uranium .............................. 39

5-14 Routine uranium urinalysis detection levels .......................................................................... 41

5-15 Factors to convert isotopic activities to total uranium ............................................................ 43

5-16 Characteristics of NU (Tuballoy) ........................................................................................... 43

5-17 Routine 210Po urinalysis detection levels ............................................................................... 45

5-18 Routine whole-body counting detection levels ...................................................................... 47

5-19 Mean body burdens of 137Cs from fallout in the United States............................................... 50

5-20 LC and MDA values for lung counting.................................................................................... 50

5-21 Hypothetical intakes for polonium and uranium, 1943 to 1946 .............................................. 56

5-22 Radionuclides for which reconstruction is not feasible .......................................................... 56

5-23 Actinium-227 dose assignment recommendations................................................................ 57

5-24 Americium dose assignment recommendations.................................................................... 58

5-25 Curium-244 dose assignment recommendations .................................................................. 59

5-26 Neptunium-237 dose assignment recommendations ............................................................ 59

5-27 Protactinium-231 dose assignment recommendations.......................................................... 60

5-28 Thorium-230 dose assignment recommendations................................................................. 60

5-29 Thorium-232 dose assignment recommendations................................................................. 61

5-30 MFP and MAP dose assignment recommendations.............................................................. 61

A-1 Identification match...............................................................................................................78

A-2 Assessment method codes for bioassay reports................................................................... 78

A-3 In vitro collection requirements ............................................................................................. 78

A-4 In vitro collection protocols.................................................................................................... 79

A-5 Sample units.........................................................................................................................79

A-6 In vivo detector systems ....................................................................................................... 79

A-7 In vivo collection protocols .................................................................................................... 80

A-8 Current and historical locations and default respiratory absorption type................................ 81

A-9 Zia employee access to plutonium areas .............................................................................. 94

A-10 Historical in vitro bioassay .................................................................................................... 95

A-11 Current in vitro bioassay performed, 1997 ............................................................................ 95

A-12 Routine sampling procedure ................................................................................................. 96

A-13 Plutonium urinalysis–sensitivities and analytical techniques ................................................. 97

A-14 Example urine bioassay results ............................................................................................ 98

A-15 Routine plutonium sampling frequency ............................................................................... 100

A-16 Routine sampling frequency 241Am exposures .................................................................... 102

A-17 Sampling protocol for tritium ............................................................................................... 102

A-18 Uranium routine sampling protocol ..................................................................................... 102
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 7 of 150

A-19 Sampling protocol for accidental plutonium exposures, 1967.............................................. 103

A-20 Sampling protocol for accidental plutonium exposures, 1973.............................................. 105

A-21 Sampling protocol for accidental plutonium exposures, 1998.............................................. 106

A-22 Sampling protocol for accidental 241Am exposures.............................................................. 107

A-23 Uranium nonroutine sampling protocol................................................................................ 107

A-24 In vitro bioassay (except plutonium and americium) sensitivity ........................................... 108

A-25 Whole-body counting detection levels................................................................................. 109

A-26 LC and MDA values for lung counting.................................................................................. 111

A-27 In vivo bioassay routine frequency...................................................................................... 112

A-28 Amount of plutonium administered to subjects by intravenous injection 

and the date of the administration....................................................................................... 113

B-1 Airborne concentrations from selected buildings................................................................. 114

B-2 MACs for selected areas..................................................................................................... 117

B-3 MACs and ALs for 1964...................................................................................................... 118

C-1 Specific activity of thorium isotopes .................................................................................... 123

D-1 Respiratory protection apparatus........................................................................................ 124

E-1 Programmatic items potentially affecting the calculation or assignment of 

dose from actinides............................................................................................................. 126

E-2 Programmatic items potentially affecting the calculation or assignment of 

dose from actinides............................................................................................................. 130

E-3 Programmatic items potentially affecting the calculation or assignment of 

dose from all radionuclides ................................................................................................. 131

E-4 Programmatic items potentially affecting the calculation or assignment of 

dose from ”alpha emitters”.................................................................................................. 137

E-5 Programmatic items potentially affecting the calculation or assignment of 

dose from unknown radionuclides....................................................................................... 138

E-6 Programmatic items potentially affecting the calculation or assignment of 

dose from plutonium ........................................................................................................... 139

E-7 Programmatic items potentially affecting the calculation or assignment of 

dose from plutonium and/or uranium .................................................................................. 139

E-8 Programmatic items potentially affecting the calculation or assignment of 

dose from thorium............................................................................................................... 140

E-9 Programmatic items potentially affecting the calculation or assignment of 

dose from plutonium and/or uranium .................................................................................. 141

F-1 Reported exposure incidents and results ............................................................................ 145

F-2 Summary of February 10, 1977, plutonium incident............................................................ 147

G-1 Levels of contact for certain job categories ......................................................................... 149

LIST OF FIGURES

FIGURE TITLE

5-1 Summary flow chart of LANL internal dose reconstruction process for 

period after 1975 ..................................................................................................................

PAGE

22
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 8 of 150

ACRONYMS AND ABBREVIATIONS

25 235U or the standard ~93% enrichment (LANL designation)

28 238U or depleted uranium (LANL designation)

49 239Pu (LANL designation)

AEC U.S. Atomic Energy Commission

AL action level

ALARA as low as is reasonably achievable

AMAD activity median aerodynamic diameter

BEST Bioassay Enrollment Scheduling and Tracking (system) 

Bq becquerel

CAM continuous air monitor

Ci curie

cm centimeter

cpm counts per minute

CMR Chemistry and Metallurgy Research

d day

D-38 depleted uranium mixture

DA delayed action

DAC derived air concentration

DIL derived investigation level

DL detection limit

DNAA delayed neutron activation analysis

DOE U.S. Department of Energy

DOL U.S. Department of Labor

dpm disintegrations per minute

dps disintegrations per second

DU depleted uranium (less than 0.72% 235U)

EEOICPA Energy Employees Occupational Illness Compensation Program Act of 2000

ESH Environmental, Safety, and Health

EU enriched uranium (more than 0.72% 235U)

g gram

GeLi lithium-drifted germanium

GM Geiger-Müller

GSD geometric standard deviation

HEPA high-efficiency particulate air

HPGe hyper-pure germanium

hr hour

HRL Health Research Laboratory

HT elemental tritium (tritiated gas)

HTO tritium oxide (water or water vapor)

IBF Ion Beam Facility

ICRP International Commission on Radiological Protection

IMBA Integrated Modules for Bioassay Analysis
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 9 of 150

in. inch

IVML In Vivo Measurements Laboratory (system) 

keV kiloelectron-volt, 1,000 electron-volts

L liter

LC critical level

LABDR LANL Bioassay Data Repository

LAMPF Los Alamos Meson Physics Facility

LASL Los Alamos Scientific Laboratory (until January 1981)

LANL Los Alamos National Laboratory (1981 to present)

LSC liquid scintillation counting

MAC maximum allowable concentration

MAPs mixed activation products 

mBq millibecquerel

mCi millicurie

MDA minimum detectable activity

MDTA minimum detectable true activity

MeV megaelectron-volt, 1 million electron-volts

MFPs mixed fission products 

mg milligram

min minute

mo month

MPBB maximum permissible body burden

MPC maximum permissible concentration

MPL maximum permissible level

mR milliroentgen

mrem millirem

mrep millirep

MSMA minimum significant measured activity

MT metal tritide

nCi nanocurie

NCRP National Council on Radiation Protection and Measurements

NDA no detectable activity

NIOSH National Institute for Occupational Safety and Health

NRS nonroutinely sampled

NTA neutron track analysis

NU natural uranium

OBT organically bound tritium

Or-93 Oralloy (Oak Ridge alloy); uranium enriched to 40% or 93%

ORAU Oak Ridge Associated Universities

ORNL Oak Ridge National Laboratory

OTLBD ORAU Team LANL Bioassay Database

OWR Omega West Reactor

PA prompt action

pCi picocurie

PDF Portable Document Format

PF Plutonium Facility
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 10 of 150

PHA pulse height analysis

POC probability of causation

PUQFUA plutonium body burden (Q) from urine analysis

R roentgen

RaLa radioactive lanthanum

RAS radiometric alpha spectroscopy

RCT radiological control technician

RIR Radiological Incident Report

s second

SEC Special Exposure Cohort

SRDB Ref ID Site Research Database Reference Identification (number)

SOP standard operating procedure

TBD technical basis document

TA Technical Area

TIMS thermal ionization mass spectroscopy

TRU transuranic

U.S.C. United States Code

UF uranium fluorophotometric (analysis technique)

UNAA uranium analysis using DNAA

UR uranium radiometric (extraction chemistry with radiometric alpha proportional counting)

VFPs volatile fission products

UPPU you pee plutonium (Club)

WGPu weapons-grade plutonium

wk week

X10 plutonium (LANL designation)

yr year

ZnS zinc sulfide

μBq microbecquerel

μCi microcurie

μg microgram

μm micrometer

§ section or sections
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 11 of 150

5.1 INTRODUCTION

Technical basis documents (TBDs) and site profile documents are not official determinations made by 

the National Institute for Occupational Safety and Health (NIOSH) but are rather general working

documents that provide historic background information and guidance to assist in the preparation of 

dose reconstructions at particular sites or categories of sites. They will be revised in the event 

additional relevant information is obtained about the affected site(s). These documents may be used to 

assist NIOSH staff in the completion of the individual work required for each dose reconstruction.

In this document the word “facility” is used as a general term for an area, building, or group of buildings 

that served a specific purpose at a site. It does not necessarily connote an “atomic weapons employer 

facility” or a “Department of Energy [DOE] facility” as defined in the Energy Employees Occupational 

Illness Compensation Program Act [EEOICPA; 42 U.S.C. § 7384l(5) and (12)]. EEOICPA defines a 

DOE facility as “any building, structure, or premise, including the grounds upon which such building, 

structure, or premise is located ... in which operations are, or have been, conducted by, or on behalf of, 

the Department of Energy (except for buildings, structures, premises, grounds, or operations ... 

pertaining to the Naval Nuclear Propulsion Program)” [42 U.S.C. § 7384l(12)]. Accordingly, except for 

the exclusion for the Naval Nuclear Propulsion Program noted above, any facility that performs or 

performed DOE operations of any nature whatsoever is a DOE facility encompassed by EEOICPA.

For employees of DOE or its contractors with cancer, the DOE facility definition only determines 

eligibility for a dose reconstruction, which is a prerequisite to a compensation decision (except for 

members of the Special Exposure Cohort). The compensation decision for cancer claimants is based 

on a section of the statute entitled “Exposure in the Performance of Duty.” That provision [42 U.S.C. § 

7384n(b)] says that an individual with cancer “shall be determined to have sustained that cancer in the 

performance of duty for purposes of the compensation program if, and only if, the cancer ... was at 

least as likely as not related to employment at the facility [where the employee worked], as determined 

in accordance with the POC [probability of causation1

The statute also includes a definition of a DOE facility that excludes “buildings, structures, premises, 

grounds, or operations covered by Executive Order No. 12344, dated February 1, 1982 (42 U.S.C. 

7158 note), pertaining to the Naval Nuclear Propulsion Program” [42 U.S.C. § 7384l(12)]. While this 

definition excludes Naval Nuclear Propulsion Facilities from being covered under the Act, the section of 

EEOICPA that deals with the compensation decision for covered employees with cancer [i.e., 42 U.S.C. 

§ 7384n(b), entitled “Exposure in the Performance of Duty”] does not contain such an exclusion. 

Therefore, the statute requires NIOSH to include all occupationally-derived radiation exposures at 

covered facilities in its dose reconstructions for employees at DOE facilities, including radiation 

exposures related to the Naval Nuclear Propulsion Program. As a result, all internal and external 

occupational radiation exposures are considered valid for inclusion in a dose reconstruction. No efforts 

are made to determine the eligibility of any fraction of total measured exposure for inclusion in dose 

reconstruction. NIOSH, however, does not consider the following exposures to be occupationally 

derived (NIOSH 2007a):

] guidelines established under subsection (c) ...” 

[42 U.S.C. § 7384n(b)]. Neither the statute nor the probability of causation guidelines (nor the dose 

reconstruction regulation, 42 C.F.R. Pt. 82) define “performance of duty” for DOE employees with a 

covered cancer or restrict the “duty” to nuclear weapons work (NIOSH 2007a).

• Background radiation, including radiation from naturally occurring radon present in conventional 

structures

• Radiation from X-rays received in the diagnosis of injuries or illnesses or for therapeutic reasons

 

1 The U.S. Department of Labor (DOL) is ultimately responsible under the EEOICPA for determining the POC. 
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 12 of 150

5.1.1

The purpose of this TBD is to provide information about occupational internal dose at the Los Alamos 

National Laboratory (LANL).

Purpose

2

5.1.2

 Occupational internal dose is the dose that was received by an individual 

from an intake of radioactive material while performing tasks in LANL buildings and structures or from 

activities outside buildings, such as burial of waste and monitoring of tests, where intakes of radioactive 

material could occur. 

This document contains information for reconstruction of occupational internal doses at LANL facilities 

throughout its history. LANL has a long history of research and production missions, and there is 

potential for exposures to a diverse set of radioactive materials. Early controls on the use, handling, 

and storage of radioactive material were based on maintenance of intakes below an assumed safe 

exposure or tolerance level, so measurements and recordkeeping practices might not support accurate 

dose reconstructions in all cases. For these reasons and others (NIOSH 2006, 2007b), a class of LANL 

employees was added to the Special Exposure Cohort (SEC) as described below. The addition of this 

class of employees has limited the scope of potential internal dose reconstructions at LANL to the 

following.

Scope

 Internal dose from tritium from 1950 to the present; 

 Internal dose from polonium from 1944 to 1956;

 Internal dose from plutonium from 1944 to the present; 

 Internal dose from uranium from 1943 to the present; 

 Internal dose from other radionuclides present at LANL may be reconstructed only from 1976 to 

the present. 

The remainder of this section describes the SEC, certain historical events that are important to internal 

dose reconstruction, and the overall approach to internal dose reconstruction for LANL workers. 

Sections 5.2 and 5.3 discuss in vitro and in vivo bioassay, respectively. Section 5.4 describes 

interferences and uncertainties, and Section 5.5 discusses the treatment of unmonitored intakes. 

Attributions and annotations, indicated by bracketed callouts and used to identify the source, 

justification, or clarification of the associated information, are presented in Section 5.6. 

Attachment A provides materials for reconstruction of occupational internal dose for monitored workers. 

Attachment B lists air concentrations in selected buildings. Attachment C discusses bioassay for 

radionuclides other than primary radionuclides, and Attachment D describes the respiratory protection 

program. Attachment E discusses programmatic issues to be reviewed for evaluating for potential 

unmonitored intakes. Attachment F contains descriptions of some incidents that resulted in internal 

doses, and Attachment G describes scaling of coworker dose intakes consistent with type and duration 

of exposures.

 

2

 The facility was known as the Los Alamos Scientific Laboratory (LASL) until January 1981, when the name was changed to 

Los Alamos National Laboratory. For convenience, this TBD uses LANL for all years of operation.
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 13 of 150

5.1.3

Two classes of LANL employees have been added to the SEC based on the findings and 

recommendations of NIOSH and the Advisory Board on Radiation and Worker Health Board (NIOSH 

2006, 2007b). The Secretary of the U.S. Department of Health and Human Services has defined one 

class as employees of DOE-predecessor agencies and their contractors or subcontractors who were 

monitored or should have been monitored for exposure to ionizing radiation during radioactive 

lanthanum (RaLa) operations at Technical Area (TA)-10 (Bayo Canyon Site), TA-35 (Ten Site), and 

Buildings H, Sigma, and U (in TA-1) for a number of workdays aggregating at least 250 workdays 

during the period from September 1, 1944, through July 18, 1963, or in combination with workdays 

within the parameters that have been established for one or more other classes of employees in the 

SEC. 

Classes of Employees Added to the Special Exposure Cohort

The Secretary has defined a second class of employees as all employees of DOE, its predecessor 

agencies, and DOE contractors or subcontractors who were monitored, or should have been monitored, 

for radiological exposures while working in operational TAs with a history of radioactive material use at 

LANL for an aggregate of at least 250 workdays during the period from March 15, 1943, through 

December 31, 1975, or in combination with workdays within the parameters that have been established 

for one or more other classes of employees in the SEC.

NIOSH has determined that it does not have access to sufficient information—including internal 

personnel dosimetry, workplace monitoring data, or sufficient process and radiological source 

information—that would allow it to estimate with sufficient accuracy the potential internal doses workers 

might have received. NIOSH has determined that there is insufficient information either to estimate the 

maximum radiation dose for every type of cancer for which radiation doses are reconstructed that could 

have been incurred under plausible circumstances by any member of the class or to estimate the 

radiation doses of members of the class more precisely than a maximum dose estimate. NIOSH has 

determined that it is possible to reconstruct or bound external doses, occupational medical doses, and 

the above-listed occupational internal doses.

5.1.4

When operations began at LANL in 1943, the only method of monitoring intake was through loose 

contamination swipes. Swipes with a lightly oiled filter paper were taken of surface areas likely to be 

contaminated. Any swiped area with an activity of more than 500 cpm (1,000 dpm) alpha required 

decontamination. (The efficiency of the stationary counters used to count the swipes approached 

50%.) In addition, nasal swipes (also called nose counts) were used to indicate potential intakes. 

Nasal swipes with alpha activity higher than 50 cpm indicated the need for follow-up bioassay. The first 

air samplers became available in the fall of 1944. However, the swipe technique continued as the 

primary method of detection in many areas until the early 1950s. Respiratory protection equipment 

(e.g., assault gas masks or respirators) was used as early as 1944 (Oppenheimer 1944).

Background of Internal Exposures

In 1943, radiation hazards of the project were limited to external radiation from the cyclotron, the Van 

de Graaff accelerator, radium sources, and a few micrograms of plutonium that arrived in the summer 

of 1943. In the spring of 1944, the first milligram quantities of plutonium arrived at LANL. Until that 

time, the laboratory was in the construction phase. There were also internal and external radiation 

hazards from uranium (ENSR 2002).

In 1944, the radiological hazards of plutonium had been recognized, although it was not yet realized 

that plutonium was more hazardous than radium. Safety regulations based on experience with radium-
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 14 of 150

dial paint plants were established. Measures to control personnel exposures included multiple changes 

of clothing, showers before leaving the building, use of surgical gloves and respirators, and use of 

closed systems whenever possible. These measures were primitive by current standards. Most 

workers cooperated with safety rules to the best of their ability, but the potential for contamination and 

intakes was present. During the tension and feverish activity of the development of the first atomic 

bomb, it was difficult to avoid some shortcuts in the observation and enforcement of safety rules 

(Hempelmann, Richmond, and Voelz 1973). The Laboratory went from handling a few micrograms of 

plutonium in 1943 to kilogram quantities in 1945. This provided little time for the usual development of 

safe methods of handling and safety equipment design (Schulte and Meyer 1957). As research 

determined that plutonium was more hazardous than radium, tolerance levels and maximum 

permissible body burdens (MPBBs) were reduced significantly. The MPBB for plutonium began at 5 μg 

(0.33 μCi) in 1943. This was reduced to 1 μg (0.07 μCi) in October 1944 and further to 0.6 μg (0.04 

μCi) in 1951 (Langham et al. 1962).

5.1.4.1 Early Research with Humans

The Plutonium Experiment, which involved plutonium tracers, was undertaken between April 1945 and 

July 1947. This was a joint project between LANL and the Atomic Energy Project of the University of

Rochester School of Medicine and Dentistry. If references to involvement in human experimentation 

are found in the dosimetry records or the telephone interview and are determined to be relevant to the 

dose reconstruction, then details for this experimentation are available in Section A.15.

5.1.4.2 Early Safety Efforts

Early safety efforts were based on working in safe contamination levels (LASL 1944a). In 1952, the 

Health-Safety Rules for Building 52 included restriction of access only through locker rooms and 

release of clothing with levels less than 100,000 cpm (LASL 1952a). Smoking was permitted in various 

contaminated areas at least through 1952 (LASL 1952b). In 1954, contamination measurements were 

based on contact with a shield-open Geiger-Müller (GM) tube for beta/gamma and a Pee Wee probe of 

a 55-cm2 area for alpha. The efficiency of these portable probes is approximately 10% to 15%. Swipes 

that were counted on a fixed proportional counter had an efficiency of approximately 40% to 50%, 

although the oil on the swipes that increased the collection efficiency might have decreased the 

counting efficiency slightly. A total alpha count rate of 500 cpm on a swipe corresponded to 0.007 μg or 

0.0004 μCi of the plutonium isotopic mixture of the times (Hempelmann, Richmond, and Voelz 1973).

Tolerance limits were established as the level below which the risk of health effects was considered 

acceptable (safe) to continue work or to not take immediate action to correct the condition (LASL 1945). 

The tolerance limit for wounds was 10 cpm for alpha and 0.15 mrep/hr for beta/gamma, except wounds 

contaminated with 90Sr, for which the tolerance limit was 0.05 mrep/hr. The skin contamination 

tolerance limit was 1 mrep/hr for beta/gamma, except for 90Sr, which was 0.05 mrep/hr. The skin 

contamination tolerance limit for alpha emitters was 1,000 cpm for polonium, 500 cpm for Tuballoy or 

Oralloy, and 250 cpm for plutonium. Decontamination was required when swipe results were greater 

than 500 cpm. Floors in laboratories were mopped once or twice a day to maintain safe contamination 

levels (Hempelmann, Richmond, and Voelz 1973). Corrective actions appropriate to the situation (e.g., 

decontamination of an area, release of personnel with skin contamination) were initiated if tolerance 

limits were exceeded. Workers exceeding bioassay tolerance limits, as discussed in later sections, 

might have been restricted temporarily or permanently from working with radioactive materials 

(Kolodney 1946). Early Health Group reports indicate contamination inside many respirators, which 

indicates improper storage and handling and poor fit during use (LASL 1944b).
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 15 of 150

Kilogram quantities of plutonium began to arrive at LANL in April 1945. At that time, portable alpha 

counters, continuously operating air samplers, supplied air lines, and specially made positive pressure 

masks were available. Procedures were performed in open hoods and wooden dry boxes, which were 

the precursor to the modern glovebox. Research indicated that there was potential for work and casual 

encounters with plutonium and other radionuclides at various air concentrations or surface 

contamination levels, including levels that exceeded radiation exposure or control limits. Given that 

these operations were, in many cases, the first of their kind and that health physics practices were 

being developed and implemented at the same time plutonium processes were being brought on line, 

some level of chronic or episodic intake during this early period would be a reasonable assumption for 

any worker in the Laboratory (Hempelmann, Richmond, and Voelz 1973). The Health Safety Report, 

Chemical and Metallurgical Division, May, 1944 (LASL 1944c), discussed the contamination surveys 

from D-118 and D-119 as an example. “These data are further proof that both wet and dry material are 

airborne. They emphasize the necessity for assuming that every uncovered item in the laboratory is 

contaminated” (LASL 1944c). Therefore, exposure to some amount of airborne contamination is a 

reasonable assumption for all workers during this early era, regardless of job title, even construction 

trade workers and janitors. 

From 1943, the Health Group was responsible for the establishment of health standards, specifically for 

safe levels of exposure to radiation and to radioactive and chemical materials. The Health Group’s 

primary concern was to protect the health of Laboratory employees. Formalized Health Rules were 

established for various areas (Burke 1946; Tribby 1946a). Until mid-1951, for want of adequate staff, 

the Group accepted help for monitoring radiation-related activities from staff members in the Chemistry 

and Metallurgy Research (CMR) organization. Over the years, the Health Group evolved into the 

Health Division [and its successors, Environmental, Safety, and Health (ESH) and Health, Safety, and 

Environment Divisions], with groups in each to address health physics, medical, (industrial) safety, 

biomedical research, industrial hygiene, industrial waste treatment, and environmental studies. While 

several division and group name changes have occurred since 1943, the generic Health Physics Group 

has existed since 1951. Throughout this period, that group has had the responsibility for assigning and 

scheduling bioassay analyses for intakes of all radioactive materials. Until the late 1990s, the Industrial 

Hygiene Group performed all bioassay analyses. Since the late 1990s, bioassays have been 

performed by one of the chemistry groups. The Medical Group has treated individuals accidentally 

exposed to radiation and radioactive materials, performed physical examinations, and treated industrial 

accidents.

Over the years, many improvements have been made in monitoring, bioassay techniques, safety 

equipment, and safety procedures (Schulte and Meyer 1957). Nevertheless, the potential for monitored 

and unmonitored intakes has existed throughout the history of the site.

Nuclides with the widest historical and current application throughout the LANL facilities are:

• Tritium (3

H)

• Uranium (238U, 234U, 235U)

• Plutonium (238Pu, 239Pu, 240Pu, 241Pu and, to a lesser exposure significance, 242Pu and 244Pu)

• Polonium (210Po) (through the late 1950s) 

• Americium (241Am)

These radionuclides of primary internal dosimetric concern are listed in Laboratory reports from 1943 to 

the present. (These are referred to as primary radionuclides throughout the remainder of this TBD.) 

Other radionuclides, such as the mixed fission products (MFPs) 14C, 11C, 13N, and 75Se, are associated 

with work areas and years of operation in Attachment A, Table A-8 (Inkret et al. 1998a). The fission 
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 16 of 150

products 140La, 140Ba, 90Sr, and 89Sr that are associated with RaLa operations (September 1, 1944, 

through July 18, 1963) are also of significance. Dose reconstruction has been determined to be 

infeasible for radionuclides other than primary radionuclides and americium, including doses from RaLa 

operations, in the period before January 1, 1976. 

Work areas include plutonium facilities (238Pu or 239Pu), uranium facilities, polonium facilities, tritium 

facilities, laboratory facilities, reactors, accelerators, and others. Case files might not have specific 

information about the assigned work areas of individuals, but detailed work histories might be found in 

the documents from DOL. When information about the work location is available, Table A-8 in 

Attachment A can be used to determine the probable nuclides and inhalation absorption type. 

Section 5.3 discusses in vitro methods for specific radionuclides. Excreta bioassay methods for 

determining internal exposures were developed for plutonium in late 1944 (fully implemented in April 

1945), for polonium in 1944, for uranium in 1949, and for tritium in 1950. Only workers with a significant 

potential for exposure were monitored. A survey in 1986 estimated that approximately 350 persons had 

known burdens of plutonium.

Starting in 1944, blood tests were performed after potential exposures. These were typically performed 

for blood count parameters in relation to external radiation exposure or the probability of poisoning 

rather than the concentration of radioactive material in the blood (Kolodney 1946); the results are not 

generally applicable to internal dose calculations.

Air samples, which are identifiable with an individual’s record, were collected and analyzed beginning in 

1944 but are not routinely provided with the LANL records.

As the state of the art of radiation detection progressed, whole-body counting for fission products began

in 1955, wound counting began before 1967, and chest counting began in 1970. Section 5.3 discusses 

historical and current in vivo bioassay methods.

5.1.4.3 Bioassay Program

Before the 1970s, individuals were assigned to a bioassay program as determined by the area health 

physics monitors. LANL deemed this program sufficient to ensure that all workers who might require 

monitoring were monitored. However, instances might have occurred, especially in the early history, in 

which a person not normally assigned to radiation work was asked to participate as a substitute in a 

task that involved radiation or radioactive materials. These persons were not likely to have regularly, or 

possibly ever, participated in the bioassay program. It is possible that their participation in these tasks 

was never recorded. Indications of this type of exposure might come from claimant interviews or work 

history statements. For this reason, potential unmonitored exposures are evaluated as part of the dose 

reconstruction as discussed in Section 5.5.

In the 1970s, LANL initiated an Employee Health Physics Checklist. This checklist allowed the 

evaluation of each individual for potential internal and external exposure. Individuals were placed on a 

monitoring schedule based on this checklist. The checklist is still used and was computerized in 1998 

as the Dosimetry Enrollment System.

Programmatic problems that were described during an assessment by a Tiger Team identified a 

number of issues in relation to the use of the health physics checklist (DOE 1991). These issues are 

listed by potential exposure type in Attachment E. In the modified dose reconstruction process in this 

TBD revision (see Section 5.1.6), the dose reconstructor reviews this list for items applicable to each 
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 17 of 150

type of potential exposure. Because the items have been listed in a more useful form for the dose 

reconstructor, they are not separately listed in the main body of the TBD. 

5.1.4.3.1 Individual Bioassay Results

The Oak Ridge Associated Universities (ORAU) Team has generated a database from the LANL 

Bioassay Data Repository (known, within this document, as the LABDR) to consolidate dosimetry 

records and make these data available for use in the dose reconstructions. The repository consists of 

several applications including the Bioassay Enrollment Scheduling and Tracking (BEST), In Vivo 

Measurements Laboratory (IVML), and Radiological Incident Reports (RIR) systems, as applicable. 

Electronic data from many other sources were collected and uploaded as part of the project. The new 

database is referred to as the ORAU Team LANL Bioassay Database in this document (OTLBD). The 

OTLBD database consists of the in vitro records for only the five major analytes of dosimetric 

significance – uranium, plutonium, americium, polonium, and tritium – and the results of in vivo

bioassay. Other bioassay might have been performed for a worker but the results have not been 

captured in the database. The results will most likely not be supplied by LANL without special requests. 

Autopsy results are clearly labeled (ORAUT 2009a). 

Bioassay results were supplied to the ORAU Team in both Portable Document Format (PDF) and 

Microsoft® Excel® formats; however, dose reconstructors will use the PDF files in the case records 

system. In both types of file, the information is divided into five sections: the demographics of the 

worker, Current In Vitro Analysis Data (1990-the present), Historical In Vitro Analysis Data (pre-1990), 

In Vivo Analysis Data, and Incident Data. Specific details about each section of the report are 

discussed below in the sections on the relevant radionuclides and in Attachment A. 

A description of the bioassay data that was reviewed and used to populate the ORAU Team LANL 

bioassay database, along with the detailed validation and verification that were performed as part of 

that process, is being compiled in ORAUT-OTIB-0063, Los Alamos National Laboratory Bioassay 

Repository Database (ORAUT 2009a). As a result of this project, LANL management and the NIOSH 

Project Manager have determined that the level of verification and validation is acceptable given the 

financial constraints of the Bioassay Data Repository Project. The validation and verification process is 

described in ORAUT (2009a). 

Each PDF file has a page with the description of the database and a summary of the validation process 

of the data. A section on Data Use Instructions is included. The information in these sections is not 

repeated here because it might be modified if the format of the files is revised in the future. These 

sections in the PDF file should be reviewed for each energy employee in the event that certain 

parameters might have been modified or updated in future files. Units and codes are clearly identified. 

Units might not be consistent within a sample result. For example, the sample result might be listed in 

picocuries per 24 hours, and the minimum detectable activity (MDA) might be listed in picocuries per 

sample or aliquot. Units of activity per day and activity per sample or aliquot are not

Descriptive codes and format descriptions are listed in each PDF file. Current codes are described in 

Attachment A. Information on documented incidents might be reported in the PDF file. This section 

might have captured recorded incidents that were listed in the original in vitro bioassay plutonium and 

americium database. Other incident information might not have been captured in the database.

necessarily 

equivalent.

The report of an individual may indicate that chelation therapy has been administered. Approximately 

35 workers are known to have received chelation therapy for plutonium and americium intakes at LANL. 
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 18 of 150

Notes about chelation therapy might be found in the comments section of the PDF files. The comment 

“URINEfix” indicates a bioassay sample that is associated with chelation. Chelation therapy was listed 

as a code in the original database, but only the comments (not the codes) were transferred to the 

current database. It is assumed that every chelation event in the original database included comments, 

but this is not ensured. 

5.1.4.3.2 UPPU Club

One small but significant group of workers is the UPPU Club (intended to be pronounced as “You pee 

Pu”). This group consists of individuals who accumulated a significant plutonium body burden and who 

agreed to be monitored periodically and continue to be monitored even past the end of their 

employment at LANL. Only two members were added to the group after the initial 1951 startup. 

Membership in this group can be noted in claimant interviews. Bioassay results can be found in the 

individual’s record many years past the end of employment or past the time of potential exposure. This

group has typically been monitored at 5-year intervals (Hempelmann, Richmond, and Voelz 1973).

5.1.5

The following sections describe several types of employees for which potential unmonitored internal 

doses must be considered when reconstructing internal doses. In each case, a review must be 

performed for potential unmonitored internal doses using case information. Programmatic problems 

identified in the Tiger Team assessment (DOE 1991) must be included in the review as well. For the 

convenience of the dose reconstructor, these items are listed in checklist form as Attachment E, Table 

E-1. Unmonitored doses suggested by case-specific information or implied by programmatic problems 

in Attachment E should be assigned in accordance with the hierarchies in section 5.1.6 (and Figure 5-

1). 

Unmonitored Workers and Unreported Bioassay Results

5.1.5.1 Workers Before 1949

The potential effects of exposure to plutonium were recognized early by its discoverer, Glenn Seaborg 

(Hempelmann, Richmond, and Voelz 1973). Therefore, when the first few micrograms of plutonium 

arrived at LANL in 1943, there was awareness of some of the radiological and biological effects and 

hazards. Stringent safety measures were put in place immediately, including the use of homemade dry 

boxes when practical. “However, because of the urgency of the times, work with plutonium had to 

proceed, and improvised methods of monitoring and decontamination were unbelievably primitive by 

today’s standards” (Hempelmann, Richmond, and Voelz 1973). In 1944, bioassay techniques 

consisted of swipes of both nostrils at the end of the working day. Monitoring results were described as 

whether MPCs were exceeded, not whether there was an intake. When kilogram quantities of 

plutonium arrived at LANL in 1945, the surface swipes and nasal swipe counting (as primary 

monitoring) had been replaced in certain areas by continuously operating air samplers and portable 

alpha counters. Then, by March 1945, a urine assay method for plutonium had been developed. Body 

burden estimates were made from urine results as early as 1953.

Working conditions were described as “deplorable by present-day standards” (Hempelmann, 

Richmond, and Voelz 1973) until September 1946, when the new facility at DP Site (TA-21) was 

constructed. Before the move to TA-21, suspension of work and decontamination of the area were 

required when alpha contamination exceeded 10,000 cpm (assume about 50% counting efficiency). 

Decontamination was recommended between 2,000 and 10,000 cpm but was not required. Even 

hallways and other unrestricted areas had contamination. In 1944, shoe covers that were worn by 

secretaries and others working throughout the buildings had count rates of 2,500 to 7,500 cpm (LASL 
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 19 of 150

1944b). Contamination above 500 cpm (0.007 μg or 0.0004 μCi of the plutonium isotopic mixture) was 

the reporting level for plutonium areas. In June and July 1945, over 50% of the laboratories had areas 

that routinely exceeded the maximum removable contamination level. The potential for unmonitored 

intakes was significant in the early years (1944 to 1946) for any site worker, but bioassay was provided 

only for the most exposed workers. Because the excretion of plutonium is continuous after an intake, 

significant intakes might have been identified on later routine bioassay samples with improved 

sensitivity for workers who remained at LANL.

As bioassay sensitivities and respiratory protection equipment improved, the potential for intakes 

decreased. Due to rigorous workplace monitoring, the probability that a worker could have received a 

large intake of radioactive material that was unmonitored and unnoticed was less after 1946, although 

the probability of unmonitored small intakes was larger. Periodic reports from H Division of air samples, 

contamination incidents, and hot spots continued to identify a significant number of over-tolerance 

occurrences throughout the 1960s. Respiratory protection was available but, except in a few locations, 

was only donned when a continuous air monitor (CAM) alarmed or when airborne levels approached 

tolerance or action levels (ALs). Therefore, a potential existed for an intake before the alarm. Review 

of Summary of Radiological Incident Reports (RIR) through June 1998 indicates that, while the number 

of RIRs has decreased significantly, the potential for unmonitored small intakes continues to exist 

(Bates 1998).

In addition, instances might have occurred, especially in the early history, in which a person who was 

not normally assigned to radiation work was asked to participate as a substitute in a task that involved 

radiation or radioactive materials. These workers were not likely to have regularly, or perhaps ever, 

participated in the bioassay program or potentially had their participation in the task recorded. 

Indications of this type of exposure might come from claimant interviews.

5.1.5.2 Short-Term Workers

Short-term workers, such as summer students, persons engaged in postdoctoral work, contractors, and 

teachers, might not have fully participated in routine bioassay programs. These workers were 

monitored for internal exposure only in unusual circumstances. Near their termination dates, workers 

might have received in vivo counts. Worker might have been required to submit an initial urine sample 

or have an initial in vivo count. In these cases, review must be performed for potential unmonitored 

internal doses, and items 7 and 8 in Attachment E should be included in the review. 

5.1.5.3 Zia Company and Other Contract Service Workers

The Zia Company was the service worker contractor. Zia employees participated in a separate 

monitoring program from that for Laboratory employees. As early as 1946, directives restricted the 

assignment of maintenance workers to either DP East or West (not both) because of the difficulty of 

separating the radionuclides in urine bioassay samples and monitoring DP East workers for plutonium 

when they might have been exposed in both areas (Burke 1946; Tribby 1946a). In 1954, that restriction 

was removed because of the improvement in the bioassay techniques that would permit the specific 

radionuclides to be identified. However, although listed as a monthly requirement for hazardous areas 

as early as 1946, submission of urine samples was dependent on determination of a “high nose count” 

or an “extremely hazardous job” or at the request of the DP East Section Leader (Meyer 1954).

In 1975, it was determined that the H-1 and H-5 efforts to schedule plutonium urine samples for Zia 

employees were inadequate. Only Zia employees who were permanently assigned to DP West and the 

CMR Building were being sampled; other Zia employees in plutonium areas were not. To ensure 
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 20 of 150

adequate coverage of Zia employees, the scheduling of plutonium samples for Zia was delegated to the 

Zia Safety Office. The requirement was annual samples for those employees “required to work in Pu 

areas,” (University of California 1978a). The problem arose in that Zia management interpreted this to 

mean employees enter Pu areas for any reason,” whereas a number of individuals in these areas would 

not have had a potential for intake of plutonium. “By July 1976 it became apparent that the number of 

Zia employees being sampled for Pu was growing without bounds” (University of California 1978a), and 

exceeded the expected number of bioassay samples. In 1978, in another attempt to reduce the 

number of sampled Zia employees to 500 per year; supervisors who only performed inspections were 

eliminated from the schedule. In later years, some jobs were exempt from monitoring. Attachment A 

contains a list of criteria and exempt job categories (University of California 1978a). In more recent 

years, other service contractors have participated in site activities.

5.1.5.4 RaLa Project Operations, 1944 to 1963

Workers on the RaLa Project from September 1, 1944, through July 18, 1963, might or might not have 

been monitored for potential intakes of 90Sr, 89Sr, 140La, and 140Ba; although it appears that RaLa 

workers at TA-10 (Bayo Canyon Site), TA-35 (Ten Site), and Buildings H, Sigma, and U (in TA-1) were 

monitored for external radiation. A total of about 250,000 Ci of RaLa was released to the atmosphere 

from these experiments. The highest exposures at LANL from RaLa operations were to the chemists 

who prepared the sources. 

In vitro bioassay results have not been found in claim files to date. The whole-body counting program 

was not well established until 1970, although there are reports as early as 1957 that identified specific 

nuclides for workers at the cyclotron (Buckland 1959). Neither have workplace air monitoring data for 

the RaLa areas been found to be available. Therefore, estimates of intakes and assessment of internal 

dose are not possible for RaLa operations workers (NIOSH 2006).

5.1.6

As described in the preceding portions of this introduction, doses inferred from internal dose monitoring 

records might not completely represent the total dose that was received by a LANL employee. 

Because this is the case, three types of internal exposures must be addressed in dose reconstruction 

for LANL workers: doses based on available internal dose monitoring records (measured and missed 

doses), potential unmonitored exposures, and exposures for which bioassay was or might have been 

performed but for which the records are not available. The last two effectively form a single category 

from the point of view of dose reconstruction and will be considered together in this TBD; for 

convenience, they are referred to as potential unmonitored intakes. 

Overview of Internal Dose Reconstruction Process for Los Alamos National 

Laboratory

The dose reconstructor must review the records in the case with special care to identify all potential 

internal doses, including potential unmonitored intakes. This evaluation will result in two possible 

methods for estimating internal dose: (1) calculating the dose from bioassay results (missed and fitted 

internal doses); and (2) assigning the dose. Doses are calculated from bioassay records, including the 

positive and the negative results. Doses are assigned based on the following sources: 

• Listed environmental intakes in the occupational environmental dose TBD for LANL (ORAUT 

2007a); 

• An overestimate of dose based on ORAUT-OTIB-0018, Internal Dose Overestimates for 

Facilities with Air Sampling Programs (ORAUT 2005a);
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 21 of 150

• Measured air concentrations from Attachment B of this TBD; 

• Data from incidents in Attachment F of the this TBD; or

• Coworker dose intakes in ORAUT-OTIB-0062, Internal Dosimetry Coworker Data for Los 

Alamos National Laboratory (ORAUT 2009b).

Calculation of internal doses is discussed in Sections 5.2 to 5.4, which address in vitro bioassay, in vivo 

bioassay, and interferences and uncertainties, respectively. Assignment of internal dose for potential 

unmonitored intakes is described in Section 5.5. Criteria are supplied to aid the dose reconstructor in 

determining the likelihood, severity, and duration of a potential unmonitored intake, and a methodology 

is given from ORAUT-OTIB-0062 (ORAUT 2009b) to allow the assignment of a maximum dose using 

coworker dose intakes. Due to the complexity of the dose reconstruction process at LANL and the 

change to include calculated and assigned doses, the following provides an overview. 

Figure 5-1 is a flow chart of the internal dose reconstruction process for periods after 1975. In general, 

the process follows the steps for one of three cases:

• Case 1, internal dose monitoring records are present: 

– Dose is calculated based on internal dose monitoring records in accordance with the various 

sections of this TBD, and 

– If potential unmonitored intakes are suggested by case information, the procedure for Case 

3 is 

the case is further evaluated for potential unmonitored intakes.

also

• Case 2, no internal dose monitoring records are present:

followed

– If the worker is a ZIA employee or a short-term or summer employee or

– If 

other case-specific 

information suggests a presumptive exposure, then the potential for an unmonitored intake 

is evaluated as Case 3.

none

• Case 3, case-specific information suggests the potential for an unmonitored intake (see 

additional guidance in Section 5.5):

of the above conditions applies, a potential unmonitored intake is not presumed. 

Rather, the environmental internal dose is assigned from section 4 of the LANL site profile, 

and the internal portion of the dose reconstruction is complete.

– If the intake is presumed to be plutonium, uranium, americium, tritium, or cesium, intakes of 

these radionuclides are based on the coworker dose intakes from ORAUT-OTIB-0062 

(ORAUT 2009b), 

– If the intake is of one of LANL radionuclides other than primary radionuclides: 

or

○ A presumed intake is based on the calculated coworker dose intakes for 239Pu (except 238Pu 

for 244Cm, and uranium for 232Th), 

○ The dose reconstructor scales the intake based on the methodology in Attachment G and 

doses are calculated based on the modified intake, 

and

and
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 22 of 150

The environmental internal dose is added if it is greater than 0.001 rem in any year.

 Figure 5-1. Summary flow chart of LANL internal dose reconstruction process for periods after 1975. 
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 23 of 150

5.2 IN VITRO BIOASSAY MINIMUM DETECTABLE AMOUNTS, COUNTING METHODS, AND 

REPORTING PROTOCOLS FOR PRIMARY RADIONUCLIDES

Historically, the in vitro bioassay program included the nuclides and techniques listed in Tables 5-1 and 

5-2. This section contains a detailed discussion of the analysis techniques that were used for selected 

radionuclides. Table A-24, Attachment A, summarizes in vitro bioassay sensitivities by technique and 

year. However, without a special request, only the results for the primary radionuclides (tritium, 

uranium, plutonium, 210Po, and 241Am as listed in Table 5-1) will be supplied for the worker. If a 

potential exposure to radionuclides other than the primary radionuclides (Table 5-2), or if an exposure 

to primary radionuclides that is not monitored by in vitro bioassay is suspected, dose is assigned based 

on Section 5.5. In some cases, special inquiries can be made to determine if additional bioassay 

results exist for the worker (e.g., MFPs or 90Sr). 

Table 5-1. Historical in vitro bioassay, primary radionuclides. 

Material Start year Comments

Tritium 1950

Uranium 1949 In practice, 1951

Plutonium 1944

Polonium-210 1944

Americium-241 1954

Table 5-2. Historical in vitro bioassay not available in claim 

records. 

Gross beta 1952 (maybe 1947) Not done regularly

Protactinium-231 1958 Not done regularly

Radium-226 1958 Not done regularly

Thorium-230 1958 Not done regularly

Actinium-227 1954 Not done regularly

Strontium-90 Unknown Not done regularly

5.2.1

The most serious intakes at LANL have involved isotopes of plutonium and 241Am. Operations with 

small quantities of plutonium and curie quantities of 210Po were conducted primarily in D Building in 

1944. The first urinalysis performed for evidence of plutonium uptake was performed on July 18, 1944, 

using a chemical procedure developed at the University of Chicago Metallurgical Laboratory in June 

1944. However, Health Group personnel soon discovered that the procedure, which used a 50-cm3

aliquot of a 24-hour urine sample, did not have the sensitivity to meet the tolerance level of a 5-μg 

(0.33-μCi) burden of plutonium. One of the first recorded accidents in which a human was subjected to 

a possible intake of plutonium occurred in May 1944 (Hempelmann, Richmond, and Voelz 1973). A 

minor chemical exposure on August 19, 1944, was the first recorded accident in which a human was 

subject to a possible intake of more than the 5-μg plutonium tolerance limit with the release of 10 g of 

plutonium into the face of a chemist. This accident resulted in LANL being authorized to proceed to 

develop a more sensitive procedure. The resultant cupferron extraction procedure and subsequent 

procedures are discussed later in this section (Moss 1990).

Plutonium

References to products called X10 and 49 indicate plutonium or isotopic mixtures of plutonium that are 

predominantly 239Pu (Lawrence 2004). (The “49” is a shorthand reference that is a combination of the 

last digit of the atomic mass number, 94 for plutonium, and the last digit of the atomic weight, 239 for 

239Pu.)
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 24 of 150

The chemical forms of plutonium that have been currently and historically encountered at LANL include 

oxide, nitrate, fluoride, and metal. Plutonium oxide, is the most common form; however, dose 

reconstructors typically assign the absorption type resulting in the highest dose, in lieu of detailed 

information specific to the case. When such information is not present, Types M and S are evaluated, 

and, in accordance with the technical information bulletin ‘Estimating Doses for Plutonium Strongly 

Retained in the Lung,’ (ORAUT 2008) correction factors for certain organs may result in additional 

dose. 

5.2.1.1 Respiratory Tract Absorption Type

The absorption type assigned for plutonium isotopes is dependent on the chemical process. Table 5-3 

lists absorption types in relation to selected processes (Hempelmann, Richmond, and Voelz 1973). 

Table A-8 in Attachment A provides information on absorption types associated with locations.

Table 5-3. Plutonium respiratory tract absorption type.

Process/area Isotope Compound Type Comments

Recovery area Pu-239 Plutonium nitrate M In addition, the area might have 

oxalates, oxides, hydroxide, and 

others based on source of input.

Purification Pu-239 Nitrate, oxalate, sodium 

plutonyl acetate

M

Fluorination Pu-239 Oxalate, fluoride, oxide M-S Powder dust, respirators worn.

Reduction Pu-239 Metallic, fluoride, chloride S Welding.

Heat source production Pu-238 High-fired oxides M-S See Cheng et al. 2004.

Pulmonary absorption type should be assumed according to work location or material composition, if 

known.

The analysis of 238Pu oxide particles from the CAM filters in place during a release in a room at the heat 

source facility in 1978 determined that the particles of 238Pu exhibited kinetics between Type M and S 

(Cheng et al. 2004).

A special case of excretion kinetics has been observed at LANL. The 1971 Wing 9 incident involved

intakes of high-fired 238Pu oxide that exhibited a very low excretion rate for the first 100 days after 

intake. A gradual rise in excretion rate was observed after that period, with no additional suspected 

intake (Miller et al. 1999).

5.2.1.2 Sample Collection Procedures

At the onset of the LANL bioassay program, samples were collected in a clean area after a 

decontamination shower. However, these samples had a potential for contamination. Occasional high 

values were assumed to be an artifact of the sample and not evidence of internal exposure 

(Hempelmann, Richmond, and Voelz 1973). By the spring of 1945, a Health Pass Ward was 

established at the hospital to ensure collection of contamination-free samples (Nickson 1945). This 

procedure allowed for 2 days off the job, 1 day to clean up, and one 24-hour period spent at Los 

Alamos Hospital where the urine sample was collected. The procedure was modified in 1948 to allow 

only the 1 day off the job during the sample collection period. While a contamination-free sample was 

ensured, this procedure was later deemed extremely expensive and was eliminated in 1952 (Clark 

2005). A procedure that allowed collection of an equivalent 24-hour sample while the employee was off 

the site was initiated. Between 1952 and 1958, the samples were collected in three disposable bottles 

that were carried in a kit. The employee was to collect four voidings in the three bottles, which some 
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 25 of 150

found objectionable and led them to collect only three voidings. The kits were rarely washed and 

therefore were considered a source of potential contamination. In January 1958 the kit provided four 

disposable bottles (Clark 2005). A modified procedure to collect a 24-hour equivalent sample (two 

morning and two evening voidings) with a four-bottle disposable kit continues to be used.

Studies of special, timed spot samples have been performed to provide data that are used to apply

correction by volume and specific gravity to provide realistic results in disintegrations per minute per 

day. All sample results, except thermal ionization mass spectroscopy (TIMS) results, are expected to 

be reported in units (typically picocuries) per 24-hour day in the database. TIMS results are reported in 

both activity per 24 hour and activity per sample. No sample volume is usually provided. The current 

database provides a clear description of units for each result. If MDAs are listed for an analysis, these 

values might be listed in units of activity per sample with a sample volume. Note that activity per 

sample and activity per 24 hour are not

LANL instituted a sample validation procedure (Lawrence 1978). The Plutonium Body Burden (Q) from 

Urine Analysis (PUQFUA) programs were revised through PUQFUA4. Therefore, other validation 

protocols might have been in effect after 1978. The purpose of these programs was to perform a 

statistical test of high values to determine if these results should be invalidated.

equivalent. It appears that the 24-hour sample concentrations 

have been normalized to a 1,400-ml volume.

Attachment A, Section A.4, lists the validation procedure for plutonium urine bioassay samples. This 

procedure is more rigorous than can be implemented without computer software. However, certain 

samples in the original database were marked invalid based on these criteria. Because the codes were 

not carried over to the current database, it is expected that these designations have been removed. 

However, this practice is mentioned in the event that comparison of the results in the current database 

is made to the original database or logbooks. The dose reconstructor should not attempt to implement 

this validation procedure but should use best judgment in the evaluation of samples for inclusion in the 

intake calculations. At present, Bayesian statistics are used by LANL to determine the statistical validity 

of results. Notes on these determinations are included in the comments in the database. Indication 

that LANL considered the sample as invalid and the reason for the status might be included in the notes 

that follow the results in the PDF file. These comments might or might not be relevant to the dose 

reconstructor’s determination of the validity of the result.

5.2.1.3 Missed Intakes

Intakes of plutonium can occur from both acute and chronic exposures. Chronic exposures might not 

be identified as incidents but can still result in a measurable burden of plutonium. A body burden can 

result from chronic inhalation exposure to a low-level, plutonium-contaminated atmosphere. A study of 

autopsy tissues from LANL workers with high and low potential for exposure to plutonium has shown 

that measurable amounts of plutonium, above that expected from global fallout, could occur in 

individuals with low exposure potential (Foreman, Moss, and Langham 1960). As noted above, 

smoking was permitted in various contaminated areas at least through 1952 (LASL 1952b). 

The long-term excretion pattern of plutonium isotopes permits plutonium intakes that produced 

bioassay results below the detection threshold in the early years to become detectable as the sensitivity 

of the analysis technique improved. The date of the intake might not relate directly to the last bioassay 

result below the detection level. The intake might have occurred many years earlier (Hempelmann, 

Richmond, and Voelz 1973). 
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 26 of 150

The first urine tests for plutonium were developed in 1944 and 1945. Urinalysis was difficult and timeconsuming. Therefore, personnel with the most positive nasal swipes had the most urine bioassays 

(Hempelmann 1946; ENSR 2002).

5.2.1.4 Routine Sample Frequencies

Table A-15 in Attachment A lists the routine sampling frequencies by period. In 1978, sampling 

frequencies were reevaluated for LANL and Zia employees. To attempt to reduce the number of 

sampled Zia employees to 500 per year, supervisors who only performed inspections were eliminated 

from the schedule (University of California 1978a). Special and emergency sample frequencies are 

listed in Tables A-19 to A-21 of Attachment A. 

5.2.1.5 Sample Analysis Procedures

At first, the urine bioassay analysis procedure could not distinguish adequately between plutonium and 

polonium. Plutonium samples in 1944 were occasionally contaminated with polonium (Clark 2005). 

During this period, the total alpha results were assigned to either plutonium or polonium based on the 

individual’s work history. The procedure was modified in the fall of 1944 to extract the plutonium. The 

count time was 30 to 60 minutes (Moss 1990).

The cupferron procedure was in use through late 1949. The bismuth phosphate-lanthanum fluoride 

serial coprecipitation procedure was in use between October 1949 and January 1957. No corrections 

were made for chemical blanks or counting geometry; average chemical recovery factors (82.3% 

±19.4%) might have been used to interpret the data until 1957 (Clark 2005). The alpha proportional 

counters were in use until 1957. 

Improvements were made in the coprecipitation procedure and counting techniques until the procedure 

was replaced by an aluminum nitrate extraction with neutron track analysis (NTA) (Clark 2005) counting 

in 1957. Urine samples were radiochemically processed and electroplated, and activities were 

determined by exposure to NTA emulsions. The exposure time for this method was 10,000 minutes

with a background of 0.005 dpm. An ion-exchange technique replaced the aluminum nitrate in 1963 

(Clark 2005). Use of a zinc sulphide (ZnS) counter began in 1966. All of these methods measured 

total alpha activity from plutonium isotopes.

Radiometric alpha spectroscopy (RAS) was available for use beginning in 1967 (Clark 2005). The use 

of RAS in the analysis is indicated when both 238Pu and 239Pu are reported for the same sample. 

Isotopic fractions can be selected based on process knowledge, as described in Section 5.3.1.6. 

Plutonium-241 and 241Am were not measured by total plutonium alpha procedures, and the dose 

contribution from these nuclides should be accounted for using the mixtures listed in Tables 5-4, 5-5, 

and 5-6, unless the plutonium in the intake was known to be a “pure” isotope.

From 1967 to present-day operations, counting has been performed by RAS for specific plutonium 

isotopes. Alpha spectroscopy cannot distinguish between 239Pu and 240Pu because of the similarity of 

the alpha energies. Therefore, the results for RAS that are labeled as 239Pu are actually 239+240Pu.

In 1997, TIMS and application of the class 100 clean room were added to the analytical technique. 

Samples are now routinely analyzed for 239Pu with TIMS and for 238Pu with RAS. The sensitivity of the 

analysis for 239Pu has been reduced 40-fold by the addition of TIMS (Inkret et al. 1999a). TIMS does 

not measure 240Pu unless the activity of 239Pu is above an unspecified threshold, in which case the 
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 27 of 150

results for 240Pu would be reported separately (Lewis 2006a). Therefore, the amount of 240Pu in the 

mixture should be accounted for with the use of Table 5-4, 5-5, or 5-6 when 239Pu results are listed for 

TIMS.

Table 5-4. Activity composition of nominal 3% plutonium mixture (Kinderman et al. 1953). 

Mixture designation: Fresh 5-yr 10-yr 15-yr 20-yr 25-yr 30-yr

Years of aginga: 0 5 10 15 20 25 30

Component Specific activity in mixture (Ci/g)

Pu-238 9.85E-04 8.72E-04 8.39E-04 8.06E-04 7.75E-04 7.45E-04 7.16E-04

Pu-239b 6.01E-02 6.01E-02 6.01E-02 6.01E-02 6.00E-02 6.00E-02 6.00E-02

Pu-240 6.74E-03 6.73E-03 6.73E-03 6.73E-03 6.72E-03 6.72E-03 6.72E-03

Pu-241 1.44E-01 1.13E-01 8.92E-02 7.01E-02 5.51E-02 4.33E-02 3.40E-02

Pu-242 <1E-06 <1E-06 <1E-06 <1E-06 <1E-06 <1E-06 <1E-06

Am-241 0 1.02E-03 1.82E-03 2.44E-03 2.92E-03 3.28E-03 3.56E-03

Pu-239+240 6.7E-02 6.7E-02 6.7E-02 6.7E-02 6.7E-02 6.7E-02 6.7E-02

Pu-alpha 6.8E-02 6.8E-02 6.8E-02 6.8E-02 6.8E-02 6.7E-02 6.7E-02

Total alpha 6.8E-02 6.9E-02 6.9E-02 7.0E-02 7.0E-02 7.1E-02 7.1E-02

Component Activity ratios

Pu-239+240:Am-241 NA 66 37 27 23 20 19

Pu-239+240:Pu-238 68 77 80 83 86 90 93

Pu-239b:Pu-240 8.92 8.92 8.92 8.92 8.92 8.92 8.92

Pu-241:Pu-239+240 2.2 1.7 1.3 1.0 0.83 0.65 0.51

Pu alpha:Pu-238 69 78 81 84 87 91 94

Pu alpha:Am-241 NA 66 37 28 23 21 19

Pu-alpha:Pu-239+240 1.01 1.01 1.01 1.01 1.01 1.00 1.00

Pu-241:Pu alpha 2.1 1.7 1.3 1.0 0.82 0.64 0.50

a. Time since separation of Am-241 from the plutonium mix.

b. Use Pu-239 only for TIMS results; otherwise use Pu-239+240 for RAS or total plutonium alpha.

Table 5-5. Activity composition of reference weapons-grade 6% plutonium mixture (PNNL 2003).

Mixture designation: Fresh 5-yr 10-yr 15-yr 20-yr 25-yr 30-yr

Years of aginga: 0 5 10 15 20 25 30

Component Specific activity in mixture (Ci/g)

Pu-238 8.56E-03 8.23E-03 7.91E-03 7.60E-03 7.31E-03 7.03E-03 6.75E-03

Pu-239 5.77E-02 5.77E-02 5.77E-02 5.77E-02 5.77E-02 5.77E-02 5.77E-02

Pu-240 1.36E-02 1.36E-02 1.36E-02 1.36E-02 1.36E-02 1.36E-02 1.36E-02

Pu-241 8.24E-01 6.48E-01 5.09E-01 4.00E-01 3.15E-01 2.48E-01 1.95E-01

Pu-242 1.97E-06 1.97E-06 1.97E-06 1.97E-06 1.97E-06 1.97E-06 1.97E-06

Am-241 0 5.83E-03 1.04E-02 1.39E-02 1.66E-02 1.87E-02 2.03E-02

Pu-239+240 7.13E-02 7.13E-02 7.13E-02 7.13E-02 7.12E-02 7.12E-02 7.12E-02

Pu-alpha 7.99E-02 7.95E-02 7.92E-02 7.89E-02 7.85E-02 7.83E-02 7.80E-02

Total alpha 7.99E-02 8.53E-02 8.96E-02 9.28E-02 9.52E-02 9.70E-02 9.83E-02

Component Activity ratios

Pu-239+240:Am-241 NA 12.2 6.87 5.13 4.28 3.80 3.50

Pu-239+240:Pu-238 8.33 8.67 9.01 9.38 9.74 10.1 10.5

Pu-239:Pu-240 4.24 4.24 4.24 4.24 4.24 4.24 4.24

Pu-241:Pu-239+240 11.6 9.09 7.15 5.62 4.42 3.48 2.73

Pu alpha:Pu-239+240 1.12 1.12 1.11 1.11 1.10 1.10 1.10

Pu alpha:Pu-238 9.33 9.66 10.0 10.4 10.7 11.1 11.6

Pu alpha:Am-241 NA 13.6 7.62 5.68 4.73 4.19 3.84

Pu-241:Pu alpha 10.3 8.15 6.43 5.07 4.01 3.17 2.50

a. Time since separation of Am-241 from the plutonium mix.
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 28 of 150

Dose calculations typically performed in dose reconstructions, including those with the Integrated 

Modules for Bioassay Analysis (IMBA) computer program, assume a combination of 239+240Pu that is 

treated as 239Pu. Therefore, if the results of TIMS analysis are used for calculation of intake, the 

contribution of 240Pu must be accounted for in the calculation.

Table 5-6. Activity composition of reference fuel-grade (12%) plutonium mixture (PNNL 2003).

Mixture designation: Fresh 5-yr 10-yr 15-yr 20-yr 25-yr 30-yr

Years of aginga: 0 5 10 15 20 25 30

Component Specific activity in mixture (Ci/g)

Pu-238 1.71E-02 1.64E-02 1.58E-02 1.52E-02 1.46E-02 1.40E-02 1.35E-02

Pu-239 5.26E-02 5.26E-02 5.26E-02 5.26E-02 5.26E-02 5.26E-02 5.25E-02

Pu-240 2.72E-02 2.72E-02 2.72E-02 2.72E-02 2.72E-02 2.71E-02 2.71E-02

Pu-241 3.09E+00 2.43E+00 1.91E+00 1.50E+00 1.18E+00 9.29E-01 7.30E-01

Pu-242 3.93E-06 3.93E-06 3.93E-06 3.93E-06 3.93E-06 3.93E-06 3.93E-06

Am-241 0 2.19E-02 3.89E-02 5.22E-02 6.24E-02 7.03E-02 7.63E-02

Pu-239+240 7.98E-02 7.98E-02 7.98E-02 7.97E-02 7.97E-02 7.97E-02 7.97E-02

Pu-alpha 9.69E-02 9.62E-02 9.56E-02 9.49E-02 9.43E-02 9.37E-02 9.32E-02

Total alpha 9.69E-02 1.18E-01 1.35E-01 1.47E-01 1.57E-01 1.64E-01 1.69E-01

Component Activity ratios

Pu-239+240:Am-241 NA 3.64 2.05 1.53 1.28 1.13 1.04

Pu-239+240:Pu-238 4.67 4.86 5.05 5.24 5.46 5.69 5.90

Pu-239:Pu-240 1.93 1.93 1.93 1.93 1.93 1.93 1.93

Pu-241:Pu-239+240 3.87E+1 3.05E+1 2.40E+1 1.88E+1 1.48E+1 1.17E+1 9.16

Pu alpha:Pu-239+240 1.21 1.21 1.20 1.19 1.18 1.18 1.17

Pu alpha:Pu-238 5.67 5.87 6.05 6.24 6.46 6.69 6.90

Pu alpha:Am-241 NA 4.39 2.46 1.82 1.51 1.33 1.22

Pu-241:Pu alpha 31.9 25.3 20.0 15.8 12.5 9.91 7.83

a. Time since separation of the Am-241 from the plutonium mix.

In the current Bioassay Repository, if TIMS results are listed for a sample the 239Pu RAS analysis for 

that sample could be footnoted with an asterisk (*). The footnote states that 240Pu activity is not 

measured in the result. The comment is an artifact of the logic in the database and could be misleading 

to the dose reconstructor. The flag is applied to all results listed for the sample when TIMS analysis is 

included. RAS cannot distinguish between the energies of 239Pu and 240Pu; therefore the activity of 

240Pu, if any, is included in the RAS activity listed as “Pu-239” (Lewis 2006a).

MDAs from procedures and reports for the various radiochemical and counting methods are listed in 

Table 5-7. The MDAs that are reported with the results in the Bioassay Repository are frequently in 

units of picocuries per sample. The results of the analysis are listed as actual values (positive, 

negative, or zero) but in units of picocuries per 24-hour

Urine, fecal, and tissue samples have been analyzed for plutonium. Most in vitro bioassay samples 

have been urine. Fecal and tissue samples have been performed only if requested in special 

circumstances.

sample. These are not equivalent units. The

sample size is usually listed. The dose reconstructor should normalize the units before determining if 

the reported results are above or below the MDA. TIMS results are listed in picocuries per 24-hour

sample and picocuries per sample; however, no sample volume is typically listed and the listed results 

are not identical. Therefore, the results in picocuries per 24 hours should be used for the dose 

reconstruction. No MDA value is typically listed with the TIMS results, so the MDA in Table 5-7 should 

be used to determine if results are above the MDA.
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 29 of 150

5.2.1.6 Plutonium Isotopic Ratios (Mixtures)

Dose assessments at LANL were historically based solely on the isotopes identified in the analysis. At 

present LANL dose assessments are based on a 6% plutonium mixture. The amount of 241Am present 

(i.e., the aging) in a sample with positive 239Pu or 238Pu is determined by analysis for 241Am or by 

process knowledge. If the determination of the mixture is based on process knowledge, the amount of 

241Am could be noted in the comments. The dose reconstructor should evaluate any 241Am results for 

the intake. There could be a short interval between the plutonium and the americium samples (Lewis 

2006b).

Most plutonium mixtures at LANL after 1956 should be assumed to be nominally 6% weapons-grade 

plutonium (WGPu) and to be nominally 3% WGPu before 1957 (Lewis 2006b). Tables 5-4 and 5-5 list 

activity and weight ratios as referenced for DOE sites for 3% and 6% WGPu.

Historically, when total alpha measurements were made for plutonium, 239Pu was the isotope chosen 

based on process knowledge. Results of these total plutonium alpha measurements before 1967 are 

found in the database and listed as “Pu-239.” Since 1967, plutonium analysis has been performed with 

RAS. RAS cannot distinguish between 239Pu and 240Pu because the alpha energies are similar and 

overlapping. However, 239+240Pu and 238Pu can be resolved using RAS. Therefore, when quantities of 

pure 238Pu arrived at LANL in 1968, the analysis was capable of distinguishing between the isotopes of 

plutonium. If the results of both isotopes are not reported as positive, then exposure to a mixture of 

predominantly pure 238Pu is usually indicated when the 238Pu results are positive and the 239Pu results 

are below the MDA.

Based on the isotopic composition of the material that was released to the atmosphere in 1970 during 

an accident in TA-21 (DP West) involving “pure” 238Pu (by weight, with a specific activity of 14.1 μCi/μg)

a typical mixture consists of 80% 238Pu, 16.9% 239Pu, 2.8% 240Pu, 0.2% 241Pu, and 0.1% 242Pu (Meyer 

1970). “Pure” 238Pu was typically present in areas involving heat source technology. A typical pure 

238Pu mixture is described in Table 5-8. It is assumed that pure 238Pu would be “fresh” during 

processing in the heat source technologies. However, intakes of aged pure 238Pu could be possible 

during decontamination or decommissioning activities in areas where pure 238Pu was processed. Table

5-8 includes the activity ratios necessary to calculate the appropriate mixture for aged “pure” 238Pu. 

Potential ingrowth of 241Am can be determined from the ratios in Table 5-8. 

At LANL, the predominantly pure 238Pu has always been processed in separate dry boxes and 

gloveboxes from WGPu to prevent cross-contamination of the two forms.

No definitive historical information on the 240Pu:239Pu atom ratios of LANL sources exists, and there is 

no information on how the ratios vary with time and location. However, before 1970, plutonium at LANL 

derived from the Hanford Site (Gallaher and Efurd 2002). According to environmental impact studies, 

atom ratios from LANL plutonium that was released before 1970 would have been 0.01 to 0.03 

240Pu:239Pu. LANL plutonium derived from the Savannah River Site starting in 1970. The atom ratio of 

LANL plutonium that was released after 1970 was expected to be 0.05 to 0.07, although ratios up to 

0.13 indicated the sediment was affected by LANL plutonium (rather than deriving purely from global 

fallout). Ratios above 0.07 in environmental samples were indicative of a mixture of global fallout and 

LANL plutonium.
Document No. ORAUT-TKBS-0010 5 Revision No. 01 Effective Date: -10/15/2009 Page 30 of 150

Table 5-7. Plutonium bioassay sensitivity as listed in procedures and reports.

Nuclide Period

Sample 

typea Technique (era)

MDA level

Unit/24-hr sampleb Reporting limitUnit/24-hr sampleb Tolerance limitcUnit/24-hr sampleb

Pu–total 

alpha

1944 U Cupferrond 0.7 pCi (Clark 2005)e >0.8e pCi 7 cpm/24 hr or 6.3 pCi

1945–1949 U Cupferrond 0.16 to 0.05 pCif

(1 dpm [0.45 pCi] DL [Moss 1990])

>0.8e pCi 7 cpm/24 hr or 6.3 pCi

1949–1/1957 U Biphosphate/alpha counting 0.20 to 0.07 pCif 2 dpm or 0.9 pCi

0.4 pCig 

7 dpm/24 hr or 3 pCi

1/1957–1965 U Aluminum nitrate/NTA 0.03 pCih

(0.05 dpm at 99% confidence)i 0.2 dpm or 0.09 pCi 7 dpm/24 hr or 3 pCi

1966 U ZnS 0.03 pCij(0.07 dpm)

Pu-239 1967–1997 U RAS (PHA) only starting in 

1971

0.03 pCij(1 mBq)

1977–1981 F RAS (PHA) 1 nCi/sample (less if Am-241 ratio known)

1981–1983 F Phoswich detector, 4-cm 

sample thickness

0.4 nCi/sample or 400 pCi/sample

(17-keV X-rays)

1982–1986 U Alkaline earthi

Oxalatei

Rapidi

Alkaline earth - 0.015 pCi/800 cm3

Oxalate – 0.015 pCi/700 cm3

Rapid – 1.6 pCi/L

1997–2002 U TIMS/RAS (alpha specl) 1.03E-4 pCi/ 24-hr sensitivity

7.6 μBq/24-hrk

1997–present U RAS (alpha spec) 8.1E-3 pCi 

2003–present U TIMSl 3E-4 pCi 

Pu-238 1967–1971 U RAS (alpha PHA) 0.03 pCim 0.2 dpm/24 hr investigate

1971–1976 U RAS (alpha PHA) 0.03 pCim

1977–1997 U RAS (PHA) 0.03 pCim

1977–1981 F RAS (PHA) 0.4 nCi/sample

1981–1983 F Phoswich detector, 4-cm 

sample thickness

0.2 nCi/sample (17-keV X-rays)

1997–present U RAS (alpha spec) 8.1E-3 pCi or 300 μBq

a. U=urine, F=fecal

b. Unless otherwise noted.

c. The tolerance limit is considered the level under which it was safe.

d. “A successful method of analyzing urine was developed in Jan. 1945 but could not be used as a routine test until a contamination free laboratory (ML Building) was ready for use 

in Feb. 1945” (Hempelmann 1946).

e. Not adjusted for potential chemical recovery Clark (2005). Assuming that “0.05 dpm/sample” refers to the entire 24-hours sample.

f. Background count rate 1 cpm (changed to 0.1 cpm at some unknown time before 1957), 1,000-minute count time, 50% efficiency, average recovery 82.3% ±19.4% (1945 to 

1949) and 67% ±21% (Nov 1949 to Jan 1957).

g. Results above these values were considered high (i.e., positive) and subject to statistical investigation. Source: Lawrence (1978).

h. Campbell et al. (1972); McInroy et al. (1991).

i. Procedures on record, not considered the default analysis method. Verify method used from dosimetry monitoring record.

j. Moss et al. (1969).

k. 2x sensitivity

l. TIMS results are for Pu-239 only; the results do not include Pu-240, which is indistinguishable from Pu-239 by other types of analysis.

m. University of California (1978a).

n. Inkret et al. (1999a). TIMS with ultra-trace chemistry and class-100 clean room and alpha spectroscopy methods. Use of alpha spectroscopy allows direct measure of chemical 

efficiency and detection of Pu-238. TIMS quantifies Pu-239 only, RAS quantifies Pu-239+240.
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 31 of 150

Table 5-8. ”Pure” 238Pu mixture (Meyer 1970).

Mixture 

designation: Fresh 5 yr 10 yr 15 yr 20 yr 30 yr

Years of aginga: 0 5 10 15 20 30

Component

Weight 

fraction 

(fresh) Specific activity in mixture (Ci/g)

Pu-238 0.80 1.370E+07 1.317E+07 1.266E+07 1.217E+07 1.170E+07 1.081E+07

Pu-239 0.169 1.048E+04 1.048E+04 1.048E+04 1.048E+04 1.047E+04 1.047E+04

Pu-240 0.028 6.352E+03 6.349E+03 6.345E+03 6.342E+03 6.339E+03 6.332E+03

Pu-241 0.002 2.061E+05 1.620E+05 1.274E+05 1.001E+05 7.871E+04 4.864E+04

Pu-242 0.001 3.933E+00 3.933E+00 3.933E+00 3.932E+00 3.932E+00 3.932E+00

Am-241 NA 0 1.463E+03 2.601E+03 3.484E+03 4.167E+03 5.094E+03

Component Activity ratiosb,c

Pu-238/Pu-239 NA 1,307.3 1,256.7 1,208.0 1,161.3 1,117.5 1,032.5

Pu-238/Pu-240 NA 2,156.8 2,074.3 1,995.3 1,919.0 1,845.7 1,707.2

Pu-238/Pu-241 NA 66.5 81.3 99.4 121.6 148.6 222.2

Pu-238/Pu-242 NA 3.483E+06 3.349E+06 3.219E+06 3.095E+06 2.976E+06 2.749E+06

Pu-238/Am-241 NA 0 9002.1 4867.4 3493.1 2807.8 2122.1

a. Time since separation of Am-241 from the plutonium mix.

b. Calculate dose for Pu-239 and Pu-240 separately using these ratios.

c. “Pure” Pu-238 was not on site at LANL until after the introduction of RAS for analysis of bioassay samples. Therefore, 

the activity ratios from total alpha plutonium are not applicable.

However, Table 5-9 describes the plutonium mixture that was found in 1960 in dirt that was used to 

sandblast plutonium parts during the demolition of a plutonium filter facility. Thus, although the 

240Pu:239Pu ratios in pre-1970 plutonium mixtures might have been nominally 0.01 to 0.03, use of a 6% 

fresh mixture for the pre-1970 years is not unreasonable. However, in general the pre-1970 mixtures 

of 3% listed in Table 5-4 are suggested when isotopic ratios are not known (Gallaher and Efurd 2002). 

Mixtures of less than 6% (i.e., 3%) are appropriate to assume during the Manhattan Project era (Lewis 

2006a). Intakes of plutonium in the early history of the site should be considered fresh. Aging of 

plutonium was addressed with continual reprocessing, and although the nature of the process work at 

LANL involved research-grade plutonium mixtures that are usually considered fresh, there is 

indication that some americium refining took place at LANL. For this reason, aged plutonium mixtures 

should be assumed when plutonium intakes resulted from plutonium intakes during decommissioning, 

other sources where long-term residual contamination might have been expected (Lewis 2006b), and

when the source term is unknown. Assume 3% or 6% fresh plutonium mixture when specific

information is available in the case records (Lewis 2006a,b) to support this assumption. There was 

practically no americium present in the plutonium at LANL in 1944 or 1945 (Voelz, Grier, and 

Hempelmann 1985).

Table 5-9. Fresh plutonium mixture 

(Christensen, Garde, and Valentine 

1975). 

Isotope Weight percent

Pu-239 93.5%

Pu-240 6.0%

Pu-241 0.5%

Any 241Am that are observed in lung counts that were performed years after the intake can be 

assumed to be the result of (1) the ingrowth of 241Am from the 241Pu in the mixture over time or (2) the 

241Am in the initial plutonium mixture unless the incident report specifically indicates potential 

exposure to pure 241Am or positive urine bioassay for 241Am is found for the intake period.
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 32 of 150

Fuel-grade plutonium (12%) was not known to be routinely encountered at LANL. However, the 

Clementine (operational 1949-1952) and LAMPRE I (operational 1961-1963) reactors utilized 

plutonium fuel. Plutonium doses for individuals working in these facilities should be based on the 

application of a 12% fuel-grade mixture in accordance with Table 5-6. 

5.2.1.7 Correcting for Urinalysis Volume

Urine samples were corrected for estimated 24-hour excretion of plutonium based on sample volume 

and/or specific gravity (Gautier 1983). All results in the current Bioassay Repository database have 

clearly listed units with associated uncertainties that are reported as one standard deviation unless 

otherwise specified in the record. Some samples that were taken in response to suspected acute 

intakes might have been spot samples of less than 24-hour excretion. These sample results are 

either noted in the database or, usually, the results are normalized to 24-hour excretion. In addition, 

the MDAs might not be in the same units as the activity results. As stated above, activity per 24-hr 

sample and activity per sample are not

5.2.1.8 Calculating Picocuries per 24 Hours for Urine Bioassay Results

equivalent units. TIMS results in the ORAU Team LANL 

Bioassay Repository are typically listed in picocuries per 24 hours and picocuries per sample, with no 

listed sample size.

All plutonium bioassay results records have been processed through the current database before

reporting. However, should the need arise to calculate plutonium urine bioassay analysis results that 

are not fully processed, equations in Lawrence (1978) can be used to convert the listed units to 

picocuries per 24 hours. Method selection should be based on the era of the samples. These 

calculations could be necessary to resolve differences if claimant-supplied results differ from results in 

the database.

5.2.1.9 Detection Sensitivities, Reporting Limits, and Tolerance Limits

Detection sensitivity was not a primary concern in the early years of operation of LANL. The concern 

was whether a tolerance limit was exceeded. Results below the MDA were originally reported as 

less-than values [listed as “LX.XX,” where X.XX is the MDA or detection limit (DL)] or 0 until the 

1980s; after then actual positive and negative results were listed in the database. In the current 

Bioassay Repository database, all results from all periods have been listed as actual positive or 

negative values. When the MDA is not specified, it is considered to be 2 times the detection level (or 

uncertainty of the blank). Average chemical recoveries and matrix blanks were not routinely reported 

before 1957. Therefore, the values in the original dosimetry records for MDAs before 1957 were 

actually the results that would be reported if an analyzed urine sample had results that were in the 

upper 12% of the background data, assuming subtraction of the standard background count rate, 

application of average counting efficiencies, and chemical recoveries. These are the values that 

would have triggered evaluations of whether the value was positive; therefore, these are being treated 

as MDAs in this document (Lawrence 1978). However, values in Clark (2005) for average 

background count rates, recoveries, and count times, and the assumption that the entire 24-hour

sample was analyzed, would result in lower MDA values. Values were calculated from the referenced 

data using the MDA equation (HPS 1996). Both values are listed in Table 5-7. Bioassay results in 

the database might have accompanying MDAs if the sample volume is also listed. The MDAs might 

be listed as activity per sample rather than as activity per 24 hours; this might also be true for the 

sample results. If the listed MDA values are used, the units must be normalized to activity per

24 hours. In some cases in the historical data the Activity field value might be null or zero. In these 

cases the MDA value should be used. This MDA value is either a sample MDA, if known, or an a 

priori MDA for the counting system of that era. The Uncertainty field value is valid only when an 

Activity value is present. Zero values rather than MDA values can exist as real values, although in 
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 33 of 150

historical times zero values were often reported when no activity was counted. If the MDA column is 

null, an MDA value for that era is unknown. All MDA values in the Bioassay Repository are true 

MDAs and not critical levels (LCs).

ETY

S MDA 66 b + .43 = (5-1)

where Sb = standard deviation of the background count

E = efficiency of the counter

T = count time

Y = chemical recovery

MDA = the MDA at 95% confidence

Before the start of RAS analysis in 1967, plutonium results were analyzed as total alpha plutonium. 

These results are listed as 239Pu in the Bioassay Repository database, but should be treated as total 

alpha plutonium, representative of the combination of 238Pu and 239+240Pu. Results from TIMS after

1997 are only 239Pu; there is no contribution from 240Pu. Before use of the results in dose calculations, 

the 240Pu contribution can be added to the 239Pu and entered as 239Pu, or the dose reconstructor can 

make separate calculations for 239Pu and 240Pu.

Plutonium bioassay sensitivities as listed in procedures and reports are in Table 5-9. In addition

Table 5-9 lists reporting and tolerance limits for certain years. Tolerance limits were defined as the 

level that could be accepted as safe. Reporting limits are values above the MDA that are considered 

significant for recording or follow-up. However, the database contains the actual bioassay results 

even if the results are below the reporting level. When MDA values are available in the database, 

these values are typically listed in units of activity per sample. Except for TIMS analysis, the sample 

mass is also listed. It is assumed that the per-sample results use the listed sample mass. While this 

mass is usually a mass that is approximately the 1,400-ml volume, which is the default excretion per 

day, these are not exactly equivalent. Appropriate normalization to 1,400 ml should be made.

5.2.1.10 Validation of Samples

In the original database, sample results in a worker’s dose record might be marked as invalid because 

the results did not meet the statistical criteria for a valid sample. If results in the Bioassay Repository 

are being compared to original data (or to bioassay results in files in the original database format 

before June 2005, which might still be in claim folders) and comments on the invalidity of the sample 

are noted, the dose reconstructor should review the sample information carefully and make a decision 

on the relevance of sample results. Section A.4 in Attachment A lists the protocol followed by the 

historical LANL computer software to validate a sample result. This section is provided for information 

only to enable the dose reconstructor to understand the criteria that were used to cause a sample 

result to be marked as invalid. This method should not be used

The excretion rate after an intake of plutonium typically remains consistent or gradually rises for many 

years after intake. An example is the original group of 26 individuals who received intakes in 1944 

and 1945 and still had measurable urine bioassay results in 1991 (Voelz and Lawrence 1991). 

Attachment A, Section A.10, contains actual results (urine bioassay and nasal swipes) from an 

individual with intakes confirmed by autopsy. These results are an example of the possible variability 

of results for an individual with an actual burden of plutonium.

by the dose reconstructor in the 

validation process. The current Bioassay Repository database might list comments taken from the 

original database. These comments frequently relate to the Bayesian statistical methods currently in 

use at LANL and are not necessarily relevant to the dose reconstruction.
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 34 of 150

5.2.1.11 Sampling Protocol for Special Accidental Exposure

Sampling protocols for special and accidental exposures are listed in Attachment A. However, review 

of the current Bioassay Repository database of workers does not completely support the bioassay 

frequencies that are listed in the internal dosimetry program. An example is a sample marked as 

involving chelation. The previous sample, taken 7 months earlier, appeared to be part of the routine 

program. The next sample after the chelation was 7 months later. There was no sampling in 

between, no initial sample to confirm the intake, and no information on the date or details of the 

intake, except a notation of “URINEFIX” on the sample, which is assumed to be the code for 

chelation. 

The incident section in the PDF file might list information on an incident that relates to the results. 

These records can provide more precise information on known or suspected intake dates. In addition, 

the record of nasal swipes, positive or negative, might be available in the incident section. 

Tables A-19 to A-21 in Attachment A list the historical and current protocols for sampling for 

suspected plutonium exposures.

5.2.1.12 Nasal Swipes

Until the first urine bioassay analysis was perfected and available in February 1945, nasal swipes 

(sometimes called nose counts at LANL) were relied on to be a qualitative indicator of plutonium 

intake (Hempelmann 1946). After the development of urine bioassay techniques, nasal swipes were 

used to indicate the need for follow-up bioassay, although it was not always performed immediately 

after a positive nasal swipe (McInroy et al. 1991). The maximum permissible level (MPL) was 50 dpm 

alpha per nostril. Positive nasal swipes can aid the dose reconstructor in determination of a plausible 

intake date. Dose reconstructors should not consider the absence of activity above 50 dpm a reason 

to invalidate bioassay results because many circumstances can contribute to a negative nasal swipe if 

an intake has occurred. A less-than-MPL nasal swipe, however, might have precluded a follow-up 

bioassay. In addition, in 1944 LANL began the practice that if the two nasal counts varied significantly 

the higher nasal count was considered spurious (Hempelmann 1944d) and the lower of the two 

counts was considered the true count (LANL 1994). Information about a positive nasal swipe can 

often be found in the incident section of the PDF file for the individual.

Intakes are not calculated from nasal swipe results, but the results of nasal swipes are used as 

indicators of possible intakes. Nasal swipes continue to be used in the present bioassay program as 

an indicator of possible intakes.

5.2.1.13 Computer-Based Calculations and Sorting

Beginning in 1959, LANL routinely used several computer-based systems to track plutonium bioassay 

and intakes. The first IBM-704 plutonium body burden report was made in September 1959. This 

program estimated the body burden of the employee and the amount of increase in body burden at 

6-month intervals (Lawrence 1978). 

Of relevance to dose reconstruction, these computer programs allowed the tracking of individuals to 

ensure that those working with plutonium or frequenting plutonium areas were on the appropriate 

routine or special monitoring programs.

Details of these historical computer programs are in Section A.4 of Attachment A. 
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 35 of 150

5.2.1.14 LANL Bioassay Data Repository

All plutonium data reside in the LABDR database (also known as the Bioassay Repository). 

Plutonium results, current and historical, have been converted to units of activity (typically picocuries) 

per 24 hours. All results are reported in actual values (i.e., positive, negative, or zero). Uncertainty 

values are listed for each measurement. Codes for the interpretation of the records are in the PDF file 

that is supplied for each individual’s results. Records of chelation therapy for plutonium might be 

noted in the comment field. However, the code in the original database that indicated that the sample 

was affected by chelation was not captured when the records were transferred from the original 

repository to the LABDR. Only comments with the results were transferred. It is assumed that all 

chelation events would have comments, but that might not be the case. The standard comment 

notation is “URINEfix” with no information on intake date, etc. The dose reconstructor should carefully 

evaluate any grouping of high bioassay results. However, only approximately 35 workers have 

undergone chelation therapy in the history of the site. Therefore, if chelation therapy cannot be 

validated, chelation should not be assumed for positive bioassay results.

5.2.2

At LANL 241

Am is usually encountered as a trace contaminant in plutonium mixtures. However, there 

is also potential for exposure to pure 241Am at LANL during actinide research activities and processing 

of plutonium mixtures. A procedure for determining 241Am in urine was in development in 1948 (LASL 

1948b). No exposure to pure 241Am was likely before the beginning of the americium bioassay 

program in 1954.

Americium

All americium data reside in the LABDR database (Bioassay Repository). Americium-241 results, 

current and historical, have been converted to units of activity (typically picocuries) per 24 hours. All 

results are reported in actual values (i.e., positive, negative, or zero). Uncertainty values might be 

listed for each measurement. Codes for the interpretation of the records are in the PDF file that is 

supplied for each individual’s results. Records of chelation therapy for americium might be noted in 

the comment field, usually with the notation “URINEfix.” However, as with the plutonium data, the 

code in the original database that indicated that the sample was affected by chelation was not 

captured when the records were transferred from the original repository to the LABDR. Only 

comments with the results were transferred. It is assumed that all chelation events would have 

comments, but that might not be the case.

There is no historical indication that workers participated in the americium bioassay program only if 

there was a potential for exposure to pure americium. Therefore, plutonium mixtures should not be 

inferred from americium bioassay results if no plutonium bioassay results are found for that period. At 

present, in vivo measurements might be performed to confirm the plutonium aging if plutonium 

bioassay results are above the MDA (Lewis 2006b). However, if the selected plutonium mixture, 

based on plutonium bioassay results, indicates the presence of americium, the absence of americium 

bioassay should not preclude the calculation of the dose from the americium contribution to the 

plutonium mixture.

Before the mid-1990s, very few individuals submitted urine samples that were analyzed for 241Am. 

Samples were submitted in response to incidents that involved exposures to sources of 241Am that are 

not contained in the plutonium mixture. Estimates of internal doses are based on the results of chest 

counts and urine bioassay (Inkret et al. 1998b). Urinalysis is the principal bioassay method for 

assessment of intakes. Publication 68 from the International Commission on Radiological Protection 

(ICRP) lists the respiratory absorption type for americium as Type M (ICRP 1995). However, if 241Am 

is determined to be part of a plutonium mixture or ingrowth from the intake of a plutonium mixture, the 

absorption type for the plutonium mixture should be assumed for the 241Am.
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 36 of 150

The current routine and special sampling programs have protocols similar to those for the plutonium 

program (Inkret et al. 1998c). Table A-16 in Attachment A describes the routine sampling frequency. 

Table A-22 describes the current program of special sampling frequencies. The potential for intake 

can be indicated by a positive nasal swipe. A positive nasal swipe is indicated by 50 dpm alpha in 

either nostril. However, lack of a positive result on a nasal swipe should not eliminate the possibility 

of an intake.

Because americium was identified as a radionuclide for which dose reconstruction is not feasible from 

the beginning of LANL operations through the end of 1975, dose reconstruction is performed only for 

1976 and after.

5.2.2.1 Minimum Detectable Activities

The historical practice emphasized being below the tolerance level rather than being below the MDA. 

All bioassay results are reported in picocuries per 24-hour sample; therefore, the listed urine bioassay 

MDAs in Table 5-10 have been normalized to picocuries per 24-hour sample. MDAs in the LABDR 

are listed in units of picocuries per aliquot. However, if no volume or mass for the aliquot is given, the 

sample volume should not be considered the aliquot volume. Results below the MDA were 

historically reported as less-than values (“LX.XX”, where X.XX is the MDA) or 0 until the 1980s. In the 

current LABDR database all results from all periods have been listed as actual values, positive or 

negative. When not specified, 2 times the detection level (or uncertainty of the blank) is considered 

the MDA. Table 5-10 lists MDAs. MDAs are based on MDAs listed in procedures and reports.

Table 5-10. Americium-241 bioassay techniques and sensitivities.

Sample 

type Year Method MDA

Urine 1954–1957 Unknown 9.E-01 pCi/24 hra

Urine 1958–1982 Chemical extraction/ 

proportional counting

2.E-01 pCi/24 hrb

Fecal 1977 Phoswich 4.E-02 nCi/sample

Urine 1983c–2002d Coprecipitation/alpha spece 1.5E-02 pCi/24 hr

Fecal 1983c Am/Pu screening/phoswich 1.E-02 nCi/sample

Urine 2003–present RAS 1.0E-02 pCi/24 hr

a. No MDA available, use derived investigation level (DIL); tolerance level 3.1 pCi/sample

(LASL 1954). 

b. McClelland (1958); method can carry over thorium, plutonium, curium, actinium, and 

neptunium. Exact start or end date of this MDA is not known.

c. Gautier (1983); exact start date of the MDA is not known.

d. Sent to an offsite laboratory from 2000 to 2002.

e. Inkret et al. (1998d).

If no MDA value is provided with the sample or the MDA provided with the results cannot be 

normalized to a 24-hour volume, dose reconstructors should use the values in Table 5-10 to 

determine if a listed bioassay result is above the MDA or to calculate missed dose using detection 

levels.

5.2.3

Tritium bioassay started in 1950. The program was formalized in 1952 (Clark 2005). Tritium was 

encountered in several forms: tritium oxide as water or gas (HTO), elemental tritium or tritiated gas 

(HT), organically bound tritium (OBT), and metal tritides (MTs). Each form has unique characteristics. 

Each year approximately 100 individuals are monitored for tritium intakes at LANL. If the work 

location was in the Health Research Laboratory (HRL) and the Ion Beam Facility (IBF) Buildings, the 

predominant form of tritium was OBT. Stable metal tritides were handled in TA-35 (Building 1), and 

Tritium
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 37 of 150

the TSF (TA-21-209, Room 179). Outside these facilities, in cases where exposure to OBT or MT are 

not implied by case-specific information, assume HTO,. Guidance for estimating doses for OBT and 

SMT is available in ORAUT-OTIB-0066, Calculation of Dose from Intakes of Special Tritium 

Compounds (ORAUT 2007b).

The potential for tritium intakes at the cyclotron and Van de Graaff accelerators might have existed

before the beginning of the bioassay program in 1950. Site locations for potential exposure to tritium 

are in Table A-8 in Attachment A. Bioassay results for tritium are usually available in the Bioassay 

Repository. However, if only whole-body dose is listed in the dosimetry records and the bioassay 

results are not available, note that all dose results from 1950 to 1988 were reevaluated by J. N. P. 

Lawrence in 1989 (Clark 2005) using ICRP Publication 30 parameters (ICRP 1979). 

5.2.3.1 Organically Bound Tritium

Information on potential exposures to organically bound tritium is not complete, but locations such as 

the HRL and IBF Buildings list labeled compounds as the principal type of radionuclides in these 

facilities. Guidance on the calculation of doses from OBT is given in ORAUT-OTIB-0066, Calculation 

of Doses from Intakes of Special Tritium Compounds (ORAUT 2007b).

5.2.3.2 Metal Tritides

Tritium exposures in the form of MT aerosols were possible. The compounds include the chemical 

hydrides and dihydrides of hafnium, erbium, titanium, and zirconium, and other metals. Uranium beds 

were utilized to capture tritium, so uranium tritide was also present (Nasise 1988, page 5). 

Information on building locations and years of operation when MTs might have been encountered is 

not complete; however, two facilities, in TA-35 (Building 1), and the TSF (TA-21-209, Room 179) were 

designed especially to handle highly-reactive tritides. Glovebox lines were used to contain the tritides, 

and a recycling system was used to capture escaping tritium (Harper and Garde 1982, page 2; Nasise 

1988, page 3). Experimentation with metal tritides required dry, inert environments in order to prevent 

contamination of the material under experimentation with water from the environment. 

Though an early tritium facility (HPTL) utilized open-faced hoods rather than gloveboxes, HPTL facility 

had as inventory only HF gas and HTO (Horak 1995), and the sealed apparatus in the hoods acted as 

a containment barrier. In contrast, the facility for handling MT built in 1953 (TA-35-2) did use a dry 

glovebox system (Harper and Garde, page 2). No specific information on the form of the tritium 

processed in the Weapon Subsystem Laboratory, which used hoods that could be completely 

enclosed, has yet been located.

ORAUT (2007b) provides guidance on the evaluation of metal tritide (MT) intakes. MTs are referred 

to as tritium particulates. In addition to potential exposure in the facilities listed above, the claimant 

telephone interview can provide indications that a person was exposed to MTs. 

Interpretation of codes and units of the bioassay data for tritium is discussed in Attachment A, 

Section A.2. 

5.2.3.3 Analytical Techniques

From startup of monitoring in 1950 until 1969, HTO in urine was analyzed by hydrogen evolved by 

dropping the urine sample on calcium carbide, collecting the hydrogen in a vacuum, and drawing the 

hydrogen into a glass electrometer chamber. In 1952, a gross GM-type system replaced the glass 

electrometer. In 1954, the tolerance for tritium in urine was 250 μCi/L (McClelland and Milligan 1954). 

The analytical range was 1 to 250 μCi/L. In 1958, an internal GM counting technique was used. The 
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 38 of 150

reported MDA for this method was 1 μCi/L. In 1970, liquid scintillation counting (LSC) began and is 

still in use today. With the introduction of LSC, the DL was 1 μCi/L. The MDA consistently improved.

For tritium results, the denominator used for reporting purposes was per liter

Results of samples were initially listed as “0” or coded as “LX.XX” (where X.XX is the MDA) to indicate 

that the value was less than the sensitivity or reporting level of the analysis. In the late 1980s, the 

practice of recording the result, positive or negative, was begun, although as late as 1990 “0” was still 

being entered into bioassay results when the value was below minimum detection levels (Lawrence 

1990a). MDA values are listed in Table 5-11 and summarized in Attachment A, Table A-24. All 

results in the Bioassay Repository database are likely to be reported in actual values (positive, 

negative, or zero). 

of urine. In 1992, all 

data from 1950 through 1992 were converted to microcuries per liter. Tritium bioassay results can be 

found as early as 1946.

Table 5-11. Tritium urine bioassay sensitivity levels.

Time Detection level Reporting levelc Counting method

1950–1951 1 μCi/La Electroscope

1952–1953 1 μCi/La GM counter

1954–1957 1 μCi/L

1958–1968 1 μCi/L Internal GM

1969b–1987 0.02 μCi/Lb,c 1 μCi/L LSC

1988–1998 0.01 μCi/Lc 0.1 μCi/L LSC and 1 ml raw urine

1999–present 0.005 μCi/L LSC

a. Expected to be the same as 1954.

b. Gautier (1983).

c. Use the Reporting Level as the MDA because results below these values were not 

reported (Clark 2005).

5.2.3.4 Routine Sampling

The routine sampling protocols for tritium bioassay throughout the history of the program are listed in 

Attachment A, Table A-17. Historically, persons working with at least 1 Ci of tritium were required to 

participate in a bioassay program of sampling every 2 weeks. In 1971, the program was changed to

determine the frequency of sampling based on the level of tritium in the previous urine sample (Clark 

2005). 

In 1998 the required turnaround times for tritium samples of all priorities was final completion within 

10 days. The preliminary report was due on priority 1 and 2 samples within 3 hours or by Monday 

morning (Inkret et al 1998c). [The definitions of priorities 1 and 2 are not specified.]

5.2.4

Bioassays were initiated in 1949 with analyses for uranium mass as an indicator of exposure (Clark 

2005). Historical uranium at LANL was primarily either depleted (often referred to as D-38 or 28) or 

enriched (referred to as Or-93 or 25) with a variety of isotopic ratios (Little, Miller, and Guilmette 

2003a). Uranium exposures from 1950 to 1957 were expected to be primarily from depleted uranium 

(DU) because lathing and milling operations took place in large rooms with no exhaust systems. 

Enriched uranium (EU) work was conducted in dry boxes, so the potential for intakes was lower (Clark 

2005). Welding on DU was introduced in 1951. However, ventilation and other measures prevented 

serious overexposures (Shipman et al. 1951). Engineering controls, especially ventilation, were 

improved in the late 1960s and reduced the potential of intakes of uranium (Lewis 2006c). 

Uranium
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 39 of 150

Conversion factors differ among DOE sites because the fractional activity of the 234U differs in the 

enrichments. Therefore, 0.17% enriched DU at LANL might have a different amount of 234U than the 

same enrichment at a different DOE site, dependent on the source of the DU and the processing. 

Tables 5-12 and 5-13 describe the nominal weight composition and fractional activity for typical DU 

and EU mixtures, respectively (Lawrence 1990b). These are the typical enrichments said to be used 

at LANL. These tables should be used unless there is specific information that states that the 

enrichments differ significantly from the enrichments listed in Tables 5-12 and 5-13. It should also be 

noted that the PDF files that were provided with worker bioassay data list a factor of 3.33 × 10-7 Ci/g 

for 238U. This is the conversion factor for pure 238U only; it does not apply to the total alpha uranium 

results that are reported, especially for the uranium analysis using delayed neutron activation analysis

(DNAA) (UNAA for uranium analysis) discussed below. The 3.33 × 10-7 Ci/g factor should not be 

used for conversion of the results to DU or EU.

Table 5-12. Nominal weight composition and fractional activities for D-38 uranium

(Lawrence 1990b).

Measure U-234 U-235 U-236 U-238 Total

Weight fraction 0.00002 0.003 0.000003 0.997

Curies in 1 g D-38 1.251E-07 6.489E-09 1.942E-10 3.355E-07 4.673E-07

Isotopic specific activity (pCi/μg) 6.25E+03 2.16E+00 6.47E+01 3.37E-01 4.673E-01

Fraction of total activity 0.2677 0.0139 0.0004 0.7180

Table 5-13. Nominal weight composition and fractional activities for OR-93 uranium

(Lawrence 1990b).

Measure U-234 U-235 U-236 U-238 Total

Weight fraction 0.011 0.933 0.002 0.054

Curies in 1 g Or-93 6.879E-05 2.018E-06 1.295E-07 1.817E-08 7.096E-05

Isotopic specific activity (pCi/μg) 6.25E+03 2.16E+00 6.47E+01 3.37E-01 7.096E+01

Fraction of total activity 0.9695 0.0284 0.0018 0.0003

DU (D-38) is the most common form of uranium that has been encountered at LANL (Inkret et al. 

1998a), although natural uranium (NU, also called Tuballoy) was used in conventional weapons 

testing from 1949 to 1970. The most common chemical forms are oxides and metal. (T3O8 can refer 

to an oxide of NU.) However, LANL has always treated uranium as either solubility class D or W. 

Urine assay data suggested that “all known LANL exposures to uranium were to a relatively soluble 

form (not Class Y)” (Lawrence 1984, 1990b). The partition between Classes D and W could not be 

determined. Class W was historically used for reporting results (Lawrence 1992a) because it 

produced larger doses. 

In addition, intake was calculated by LANL based on Class D mass of uranium, for assaying in 

relation to the nephrotoxic limit for uranium. The nephrotoxic limit is not relevant to dose 

reconstruction. However, the dose reconstructor should evaluate the data to determine the 

absorption type that is most appropriate and favorable to the claimant. Statements about welding on 

uranium, both DU and eventually EU, which would have been a potential for more insoluble forms of 

uranium, are in Shipman et al. (1951).

The LANL program calculated intakes for individuals with at least one positive urine analysis result. 

The potential for intake was indicated by a positive nasal swipe (50 dpm in either nostril). Lack of a 

positive result on a nasal swipe should not eliminate the possibility of an intake, but follow-up 

bioassay was not likely to have been performed as a result of a nasal count of less than 50 dpm. 

Otherwise, dose reconstructors can use the nominal compositions in Tables 5-12 and 5-13 as default 

compositions when no other information is available on the enrichment of the uranium (mass or 
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 40 of 150

weight fraction of 235U multiplied by 100%). The values in Tables 5-12 and 5-13 are the compositions 

at LANL at least from 1970 through 1990 (Lawrence 1990b). In addition, in Lawrence (1990b) the 

presence of 236U in the isotopic mixture is indicated. The 236U is indicative of recycled uranium (RU). 

However, LANL did not process uranium except to extrude or machine it. Oak Ridge National 

Laboratory (ORNL) was the source of the uranium that was processed at LANL. 

5.2.4.1 Sampling Protocol

Historical sampling protocols are listed in Attachment A, Table A-18. The program was initiated in 

1949 with analysis for uranium mass as an indicator of exposure. Samples were collected on Fridays 

just before workers left the site to maximize the sensitivity of bioassay detection (Lawrence 1992a). 

Since 1983, spot samples have been collected as far removed from the potential exposure time as 

possible, usually Monday mornings before workers entered the work area. The sample might have 

contained the last voiding on a Sunday evening and/or the first voiding on a Monday morning if extra 

volume was necessary. This protocol ensured that the large fraction (approximately 80%) of rapidly 

excreted uranium, experienced on the first day after an intake, was excluded from the sample 

(Lawrence 1984). Beginning in July 1993, the larger volume of sample that was required for the alpha 

spectroscopy method of analysis made the Sunday evening/Monday morning collections standard 

protocol (Lawrence 1992b).

In 1992, retroactive calculations were made by Lawrence of the data beginning in 1949. Data for 

1951 were not included in these calculations because the bound Los Alamos Notebooks could not be 

found. Clark (2005) suggests that these notebooks were never found and, therefore, 1951 data are 

missing from the records. However, review of the current database indicates a number of uranium 

bioassay records for 1951 similar to the number of records in adjacent years. Therefore, it can be 

assumed that no data are missing (Buddenbaum 2006b). For enriched 235U, the conversion of data 

into the computer files and retroactive calculations started with the 1955 data. Assays that were 

identified as 238U were converted to micrograms per day with values less than the MDA or zero set to 

the minimis MDA. Assays that were identified as 235U were converted to picocuries per day with 

values less than the MDA or zero set to the minimis MDA. In the Bioassay Repository database, the 

micrograms-per-day and picocuries-per-day results have been converted to micrograms per liter and 

picocuries per liter, respectively. These results should be normalized to a 24-hour day using an 

assumed volume of 1,400 ml/d of urine.

5.2.4.2 Uranium Analysis Techniques

Several analysis techniques were employed over the history of LANL. Techniques include 

fluorophotometric (UF), ion exchange, or extraction chemistry with radiometric alpha proportional 

counting (UR), DNAA (or UNAA), and RAS. UF and UR were total uranium techniques with the 

enrichment selected because of workplace knowledge. The uranium enrichment for DNAA/UNAA 

results was also determined by workplace knowledge.

Table 5-14 summarizes the routine urinalysis detection levels for various periods. The results of urine 

bioassay for uranium have been entered in the database. The results in 1991 or before are likely to

be reported in actual values (positive, zero, or negative) (Lawrence 1992a). Results below the MDA 

might be listed as zero or blank.
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 41 of 150

Table 5-14. Routine uranium urinalysis detection levels. 

Period Method MDA DL

Reporting 

levela

1949–1967 Fluorophotometric (DU or NU) None listedb 50 μg/Lc >100 μg/L

1968–2/1976 Fluorophotometric (DU or NU) 4 μg/Lb,d U

3/1976–1978 Fluorophotometric (DU or NU) 1 μg/Ld U

1949–1954 Anion exchange/gross alpha counting 

(possibly used)

25 dpm/L >100 dpm/L

1955–1971 Extraction/alpha proportional counting 

(U-234 alphas measured)

Unknown b,e 50 dpm/Lc >100 dpm/L

3/1971–1976 Extraction/alpha proportional counting 15 dpm/Lb >100 dpm/L

6/1976–1/1977 Extraction/alpha proportional counting 10 dpm/Lb

2/1977–12/1977 Extraction/alpha proportional counting 4 pCi/L

1978–7/1982 UNAA/delayed neutron countingf

Only U-235 counted

1 μg/L DU 

0.17% enriched

7/1982–6/1991 UNAA/delayed neutron countingf

Only U-235 counted

4 pCi/L EUf

93% enriched

1982–6/1991 UNAA/delayed neutron countingf

Only U-235 counted

4 μg/L DU 

0.17% enriched

1982–1992 Anion exchange/colorimetric (used to 

confirm UNAA >30 μg/L 

1 μg/50 cm3

(not a standard 

method of analysis)

7/1991g–1998 RASh for each isotope 0.1 pCi/L 0.05 pCi/L

1998–present RASh for each isotope 0.008 pCi/L 0.004 pCi/L

(1 standard 

deviation)

a. Exceeding Reporting Levels required investigation and evaluation (Lawrence 1984).

b. Lawrence (1984).

c. Use as detection level (MDA = 2 × DL) because this level would be used to determine positive results.

d. 50 μg/L considered positive indication of NU material in the body (Dummer 1958).

e. Specific for U-235 and U-233, 50 dpm/24-hour sample considered positive indication of EU in the body (Dummer 1958).

f.

g. Results for July 1991 through December 1991 might be listed as UNAA results in the Bioassay Repository database. 

RAS values were used directly in 1992 (Lawrence 1992b).

See Section 5.3.4.3 for full explanation of the interpretation of UNAA results including minimum reporting levels.

h. Can measure U-234, U-235, and U-238. Alpha spectrometry cannot differentiate between U-233 and U-234.

5.2.4.3 Delayed Neutron Counting

DNAA (also known as UNAA in the Bioassay Repository) was the analytical technique LANL used 

between 1978 and 1992. The technique might have been implemented as early as 1962 (ORAUT 

2009a); however, UNAA was not in general use until 1978. Interpretation of the results during this 

time was dependent on identification of the radioisotopic mixture to which the person might have been 

exposed (Lawrence 1992b). Because the person’s work history might put them with D-38 one day 

and Or-93 the next, uncertainty can be introduced into the results.

The DNAA method depends on counting the number of delayed neutrons produced by fission of 235U 

after thermal neutron activation. This method also fissions 239Pu and 233U but, according to Gautier 

(1983), 239Pu is not an interference unless the concentration in the sample is greater than 300 pCi/L 

and 233U was not a significant intake potential at LANL.

The reported results of the DNAA method for EU analyses are labeled “U-235” in the records, but 

have always included the alpha activity of 234U that is about 34 times that of 235U. Results for DU

analyses are similarly labeled “U-238,” but have always been the total uranium mass based on the 

assumed isotopic mass fractions of 0.00002 234U, 0.0017 235U, and 0.9983 238U (Lawrence 1992a).
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 42 of 150

The DNAA/UNAA was designated as a screening procedure and not a quantitative measurement. 

(Although, except for the confirmatory procedure for results above 30 μg/L, there were no other listed 

quantitative procedures for that period). At the time when DNAA/UNAA was initiated, the uranium 

exposures had been reduced due to enhanced workplace ventilation units. The ventilation units drew 

off most of the airborne particulates from machining and grinding operations so the higher exposures 

in the 1950s and early 1960s were eliminated (Lewis 2006c).

In the DNAA measurement, calibration curves were made up of standards of EU (93% 235U) and DU 

(0.17% 235U) (Gautier 1983). These calibration curves were plotted (labeled) as delayed neutron 

counts vs. 235U pCi/L or 238U μg/L. For the EU plot, concentration was in terms of total activity of EU. 

For the DU plot, concentration was in terms of total mass of DU. The calibration curve was selected 

based on the type of work, and then the concentration of uranium was interpolated.

When a worker was first entered in the system, it was noted on the health physics checklist whether 

the worker worked with EU, DU, or both. This information was transferred to the sample kit label as 

"235," "238," or "235/238." Absolutely no indication of the fraction of time or work location would have 

been recorded.

If the worker had a “235” on the sample kit, the EU calibration curve was used and an activity result 

that was labeled as 235U was calculated. If the worker had “238” on the sample kit, the DU calibration 

curve was used, and the mass result that was labeled 238U was calculated. If the worker had 

“235/238” on the sample kit, both curves were used to determine results and both a 235U by activity 

result and a 238U by mass result were calculated. No fractionation

The DNAA method actually measures the amount of 235U in the sample. Only if the ratios of 238U/235U 

and 234U/235U are well known can the uranium mass and activity in the sample be accurately 

determined (Lawrence 1992a). By the DNAA method, the minimum reporting value was 4 μg/L for DU

(0.17% enrichment) and 4 pCi/L for EU (93% enriched). For D-38 (DU), a 4-μg/L sample would 

contain 0.0147 pCi/L of 235U [(4 μg/L × 0.0017 × 2.163 × 10-6 Ci/g × 1 × 1012 pCi/Ci)/(1 × 106 μg/g)]. 

The enrichment of 235U assumed for DU at this period was 0.17%. A 4-pCi/L sample of EU (93%) 

would contain approximately 0.116 pCi/L of 235U [0.116 pCi/L ÷ 0.0284 (from Table 5-13) = 4 pCi/L]. If 

the person’s actual urine sample contained 0.116 pCi/L of 235U, the reported value for EU would be 

4 pCi/L. If the analysis was designated for both EU and DU, the D-38 value would be reported as 

approximately 32 μg/L [(0.116 pCi/L ÷ 0.0147 pCi/L) × 4 μg/L] and was labeled U-238 (Lawrence 

1992b). Uncertainty about the material of exposure, whether enriched or D-38, could introduce as 

much as a factor-of-4 error in the evaluation of a worker’s dose (Lawrence 1992b).

of the results is assumed when 

both isotopes are listed. The dose reconstructor must either use the result or the MDA most favorable 

to the claimant or make a judgment on the appropriateness of either DU or EU dependent on the 

telephone interview or DOE records.

Uncertainty for this method was considered to be 96% ±6% at 5 to 26 pCi/L for EU and 93% ±8% at 4 

to 17 μg/L for DU. 

This method was eventually deemed operationally unacceptable because of long turnaround times 

and the need to identify the mixture. LANL changed to RAS techniques between July 1991 and 

January 1992. Extreme care should be used in interpreting the results of bioassays between these 

dates. Normalization might have been done to RAS results to maintain consistency with the DNAA 

results, as discussed in footnote g to Table 5-14. 
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 43 of 150

5.2.4.4 Radiometric Alpha Spectroscopy

This method consisted of a full chemical extraction process for the uranium from the urine. A tracer of 

232U was added to determine chemical recovery. The extracted and plated uranium was counted by 

RAS and the activities of 232U, 234U, 235U, and 238U were determined (Lawrence 1992a). 

Several problems were initially encountered with this method. No chemical or synthetic urine blanks 

were used to establish the MDA. The counting blank was a blank planchet, rather than a reagent 

blank, which would produce a lower background. Chemical recoveries ranged from 10% to 60%. 

Some analyses indicated isotopic ratios outside the realm of probability. The computer program 

converted all negative values to zero. All 1991 data yielded negative values for 235U, rather than the 

expected negative and positive values. In early 1992, the computer program was still not modified to 

correct for chemical blanks.

Results that are listed as RAS in the current database, which are isotopic, should be evaluated 

according to the type of enrichment that is determined for the intake. Table 5-15 provides the factors 

to be applied to the isotopic bioassay results to estimate the total uranium as either DU, NU, or EU.

Table 5-15. Factors to convert isotopic activities 

to total uranium.

Radionuclide

Activity factora

DU NU EU

U-235 NA NA 35.211

U-234 3.736 2.058 1.031

U-238 1.393 2.033 NA

a. NA indicates that the fraction of the isotope is too small 

to be used as an indicator for the given enrichment.

5.2.4.5 Potential Missed Intakes

Programmatic problems in DU areas of TA-55 are summarized in checklist form in Attachment E. The 

dose reconstructor must consider these issues when identifying a potential unmonitored intake for 

claims with employment in these areas.

5.2.4.6 Excretion of Environmental Levels of Uranium

NU from nonoccupational intakes (primarily food and water) can be excreted in urine at levels above 

the analytical MDAs for either the elemental uranium analysis or the alpha spectrometry analysis. 

The 234U:238U ratio can be used to distinguish DU from NU. Table 5-16 lists activity ratios for NU. 

Table 5-16. Characteristics of NU (Tuballoy). 

Measure U-234 U-235 U-236 U-238 Total

Weight fraction 0.0000537 0.0072 0 0.99274

Specific activity, pCi/μg 0.33367 0.01557 0.33367 0.68291a

Fraction of total activity 0.4886 0.0228 0 0.4886

a. As listed in IMBA Version 1.0.42.

A 1992 study (Little, Miller, and Guilmette 2003a) listed the average drinking water concentrations for 

the Los Alamos/White Rock/Santa Fe area as 0.015 μg/L (0.01 pCi/L or 0.00037 Bq/L). Therefore, 

with the use of 0.01 as the factor of drinking water concentration to excretion concentration (Little, 

Miller, and Guilmette 2003a,b), LANL used the urinary excretion values of 0.00015 μg/L for elemental 

analyses, 0.0001 pCi/L for 234U and 238U, and essentially anything that was detected for 235U to 
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 44 of 150

distinguish between natural background and potential occupational exposure for uranium. In some 

cases, specific information in the worker’s file indicates that the excretion was from natural sources. 

New Mexico is known for high levels of NU in the soil and ground water. However, some LANL 

workers lived in areas of particularly high NU concentrations that ranged from 0 to 4 Bq/L (108 pCi/L) 

in 1992 and up to 6 Bq/L (162 pCi/L) in 2001 (Little, Miller, and Guilmette 2003b). These areas of 

high concentration are primarily in the Española area.

Beginning in 1992, all workers who participated in the uranium bioassay program were requested to 

bring samples of their drinking water to LANL for analysis with each urine bioassay sample (at least 

once each year), the concentrations were found to vary widely even for individuals. The results of the 

analyses of these water samples are listed with the bioassay data for the individual. When water 

sample results are noted in the record, LANL multiplied the results by 0.01 to calculate the 

concentrations of uranium isotopes to be subtracted from the individual’s bioassay results before

calculation of an occupational intake.

Background excretion of uranium in feces probably varied over an even larger range than urinary 

excretion. Fecal samples were rarely obtained for potential uranium intakes; when they were, the

investigation report should discuss how the results were interpreted.

 The factor is based on the consumption of 1.1 L of water per 

day (Little, Miller, and Guilmette 2003b). When LANL assesses dose or potential occupational 

intakes, the concentration of the most closely associated water sample might be subtracted from the 

bioassay sample results rather than an average value for all the drinking water samples for the 

individual. (There is no evidence that the individual bioassay results for the individuals in the 

database have had the drinking water concentrations subtracted. Therefore, these results should be 

used as found.) Other variables, such as drinking habits, dietary components, and individual 

physiological differences, can influence the individual’s baseline excretion rate of environmental 

uranium.

5.2.5

Polonium for use in initiators was handled in quantities of 100 Ci/mo or more starting in February 1945 

and increasing to 500 Ci/mo by December 1945. Before that time, smaller quantities were handled at 

in TA-1 at H and Gamma Buildings (ENSR 2002). DP East, Buildings 151, 152, 153, and later (1949) 

155 were also used to process polonium for initiators.

Polonium

Work at LANL with 210Po ceased in 1959. After that time, polonium was encountered only in the form 

of sealed Po-Be sources. After 1959, there were two incidents in which sealed Po-Be sources broke 

open; personnel involved in the incidents submitted urine samples for bioassay. Unless a worker was 

involved in the broken source incidents, no intakes should be calculated after 1959.

Work with 210Po was of a limited scope at LANL. Therefore, missed dose for polonium should be 

assessed only if actual polonium bioassay was performed for the individual or if there is indication, 

through claimant interview, that the employee was actually exposed to polonium. Potential missed 

dose should be assessed only during the interval covered by the bioassay or work period. In 1945, 

only two persons exceeded the 1,500-cpm/24-hr sample tolerance for polonium (ENSR 2002). 

However, University of California (1977a) stated, “During the 1940s and early 1950s, 210Po exposures 

at LASL occurred with some regularity. There was no method in routine usage to determine the 

actual exposure. (Exposure was controlled from raw urine assay data.)” Dummer (1958) stated that 

the possibility of exposure should be investigated at greater than 10 dpm/L in a urine bioassay 

sample.
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 45 of 150

Hempelmann (1946) stated that, because of the ease of the analysis, polonium urine bioassay was 

performed routinely and frequently on all individuals who worked with polonium. These records are 

expected to be in the current database.

The initial urine bioassay analysis procedure could not distinguish adequately between plutonium and 

polonium. During 1944, the total alpha results were assigned to either plutonium or polonium based 

on the individual’s work history. The procedure was modified in the fall of 1944 to extract the 

plutonium. Hempelmann (1944) indicates that this modified procedure was originally developed at 

Monsanto. LASL (1945) lists the procedure and states that the recovery of polonium on the copperplated disk was expected to be greater than 80%. Gautier (1983) contains a procedure for 210Po in 

urine with an effective date of May 1, 1955, that states that the average recovery of standards was 

90% ± 3% at the 15-pCi concentration. Discussions with a retiree indicate that no correction for 

recovery was included in the calculations [1]. Therefore; 100% recovery was assumed, and all 

polonium bioassay results and bioassay results less than the MDA should be modified to include the 

10% recovery factor that has been determined to apply to samples analyzed by the Monsanto 

method [2].

Detection limits for routine urinalysis are listed in Table 5-17. A procedure is listed for 210Po in urine in

LASL (1954) and Gautier (1983). The MDA continues to later periods unless changed.

Table 5-17. Routine 210Po urinalysis detection levels. 

Period MDA Recheck

Tolerance 

limitb

1943–1952 440 dpm/24 hra

1953 50 dpm/24 hra

1954 10 dpm/L 100 dpm/L 500 dpm/L

1955 0.1 pCi/L

1958 100 dpm/L 500 dpm/L

a. LASL (1979).

b. Workers exceeding the tolerance limit were removed from 

exposure potential until the levels dropped below the 

tolerance limit.

5.2.6

LANL has always been a center for research. As such, small-scale use of various radionuclides (in 

terms of either the number of persons or activity of the source) that are not addressed above has 

occurred throughout the history of LANL. Although some documentation has been found on bioassay 

for these nuclides, the records are not provided to the dose reconstructor as part of the case-specific 

information. Instead, doses are assigned in accordance with the guidance in Section 5.5 on 

unmonitored intakes. In addition, Attachment G provides a method to scale the intakes provided to 

present a credible exposure to individuals exposed to rarely encountered radionuclides. 

Other Limited-Exposure Radionuclides

5.3 IN VIVO BIOASSAY MINIMUM DETECTABLE ACTIVITIES, ANALYTICAL METHODS, 

AND REPORTING PROTOCOLS

In vivo counting equipment and techniques were developed in the late 1950s and have been in 

routine use for measuring X-ray- and gamma-ray-emitting radionuclides since 1970, and possibly as 

early as 1960. There is some indication that some of the counts between the beginning of the 

program in 1955 (Van Dilla 1959) and the 1960s were performed for development of the program 

rather than actual suspected intakes. Counts during this period should be evaluated as closely as 

possible for validity in the dose reconstruction. No in vivo results are expected to be found in the 

Bioassay Repository before 1969.
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 46 of 150

5.3.1

The first whole-body counter to be used at LANL was the HUMCO I (Ennis 2003), which became 

operational in 1955 (the name stood for 

Whole-Body Counters

human co

The HUMCO II became operational in 1958. The HUMCO II was housed in a count room (SB-16) 

made of 7-in. pre-World War II steel. The resolution was improved, but it remained a screening 

counter. Evaluations were made using the shadow shield or NaI(Tl) crystals. The 1958 annual 

progress report for the H-1 Group discusses “significant” measurable 65Zn in cyclotron workers 

(Buckland 1959). 

unter). The counter consisted of a large double 

cylinder with a liquid scintillation fluid (possibly trichloroethylene) that filled the annular space between 

the cylinders. The scintillation fluid was viewed with an array of 5-in. photomultiplier tubes on the 

outside wall of the cylinder. The individual was placed inside the count chamber. The count rate was 

compared to the background count rate. The system typically used two energy windows: one for 40K 

(1 to 2 MeV) and one for 137Cs + 40K Compton counts between 0.5 and 0.8 MeV. The result was 

obtained by subtracting the contribution of 40K. The system was not used for photons below 100 keV. 

This system was used to screen individuals who might have been exposed to fission products at the 

reactors or in flyovers during weapons testing. It was also used to detect the bremsstrahlung from 

90Sr intakes. The energy resolution of these counters was poor. When an elevation of the 

background in a region of interest was observed, the individual was referred for screening with either 

the shadow-shield or full-shield 4- by 8–in. NaI(Tl) crystals (Healy 1970). The sensitivities of the 

NaI(Tl) crystals were approximately the same as those for the HUMCO, but the count time was 

significantly longer.

In 1970, an in vivo counter capable of measuring four separate regions of the body began operation 

(Vasilik and Aikin 1983). Twin phoswich (CsI and NaI) detectors were placed over the lungs. The two 

layers of the detector were capable of simultaneously, yet separately, monitoring chest burdens for 

10- to 250-keV photons (NaI), for plutonium and uranium isotopes and 241Am, and for 200- to 

2,000-keV photons (CsI) for a qualitative assessment of a variety of fission and activation nuclides. A 

planar hyperpure germanium (HPGe) detector monitored the region between 10 and 250 keV with 

excellent energy resolution, and could be positioned over the liver or thyroid as needed. Last, a 

lithium-drifted germanium (GeLi) detector (which was later replaced by an HPGe detector) was 

positioned under the prone subject. This detector was primarily for whole-body assessment. This 

system could both identify radionuclides and quantify the burdens. The twin phoswich detectors were 

replaced by twin three-detector arrays of HPGe detectors in 1999. In 1990, the Tiger Team 

Assessment report indicated that the phoswich detectors were not state-of-the-art and could not 

accurately measure plutonium and americium in the presence of interfering gamma emitters, such as 

137Cs (DOE 1991). The phoswich and germanium systems were operated concurrently during the 

period from 1998 to 1999. Two of the six HPGe detectors were used when a thyroid count was 

required.

5.3.1.1 Minimum Detectable Activities and Decision Levels

Operationally, an observed signal must exceed the critical level (LC), formerly referred to as the 

maximum significant measured activity (MSMA), to result in the decision that some activity has been 

detected in the lung, body, or other organ; this is also called the decision level (DL) (Vasilik et al. 

1984). The MDA, formerly the minimum detectable true activity (MDTA), is the smallest amount of 

activity that can be in the lungs or organ such that a measurement of an individual can be expected to 

imply, correctly, the presence of activity with a predetermined degree of confidence. The MDA (or 

MDTA) and LC (or MSMA) values are listed in Table 5-18 for various years of operation. The MDA 

and LC values for lung counting are summarized in Attachment A, Table A-25. The values of MDA 

and LC are calculated in accordance with the theoretical developments of Currie and of Altshuler and 
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 47 of 150

Pasternack (Vasilik et al. 1984). No information is available on MDAs for the thyroid detector or for 

131I or 125I in the whole-body count. If available, the MDAs that are listed in the records of the worker 

should be used rather than the MDAs listed in the tables. 

Whole-body and lung counting are the primary methods for determining intakes of fission and 

activation products. However, both in vitro and in vivo results might be available.

Results in the bioassay records are generally reported in nanocuries unless otherwise indicated. 

Results that are listed as NULL indicate no detectable activity (NDA) rather than the MDA. Results 

less than the LC are marked as NDA in the database. The actual DL values and counting errors might 

be available on the report of the individual, but they are not available in the IVML database until 2003. 

However, MDAs are listed with in vivo count results in the Bioassay Repository database for 241Am

and 239Pu. Since 2004 MDAs are also listed for 235U and 234Th. In addition, every count performed 

since 1969 when the program was formalized has been recorded in the database and has a “White 

Card” on file. Counts without positive results for any analyte are marked NDA. However, the results 

of counts from before 1969 were only performed in conjunction with significant radiological events. 

Records of these measurements are in logbooks, but might only indicate that measurements were 

made, without associated results. Early in vivo measurements were made in "go/no-go" fashion using 

instruments with little or no energy discrimination capability. Negative results might not have been 

recorded but only verbally communicated to the subject. Logbooks with early in vivo measurements 

results are in LANL archives on the site (Hoover 2007).

An in vivo count spectrum is not analyzed for a fission or activation product radionuclide unless a 

peak that is associated with that nuclide is visible in the spectrum. When that peak is visible, the 

suspected nuclide is added to the library and the spectrum is reanalyzed. Visual or non-library-driven 

software recognition of a peak can be subjective and not directly correlate to MDA or LC calculations, 

especially with the broad peaks that are associated with scintillation-type detectors. For whole-body 

counts, it is not reasonable to assume that a worker was exposed to or is being monitored for all 

radionuclides potentially reportable simply because an MDA was determined and listed on the report.

In general, no information is available in the reports on the assignment of respiratory absorption type 

for specific fission and activation product nuclides. Individual guidance might be available in the 

narrative on the White Card that accompanies a worker’s in vivo bioassay report. It is unlikely that the 

information on the White Card has been transferred to the remarks section of the Bioassay Repository 

database. However, if the dose reconstructor deems it appropriate, copies of the White or Beige 

cards can be requested.

RaLa operations occurred at LANL from September 1, 1944, through July 18, 1963. The internal 

dose from intakes or potential intakes of 140La and 140Ba from RaLa operations at TA-10 (Bayo 

Canyon Site), TA-35 (Ten Site), and Buildings H, Sigma, and U in TA-1 for chemists and other 

maintenance and operations workers and security personnel cannot be assessed due to the lack of a 

well-established in vivo bioassay program during that period. Some whole-body counts might have 

been performed beginning in 1955. In vivo results might be available in the Bioassay Repository.

Table 5-18. Routine whole-body counting detection levels (Vasilik et. al. 1984; 

Ennis 2003).

a

Period Nuclide LC (nCi) MDAb,c (nCi)

1955–1958d Cs-137 8

Sr-90e 30

1959–1970d Cs-137 4

Sr-90e 30

1971–1984 Be-7 0.9 1.8
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 48 of 150

Period Nuclide LC (nCi) MDAb,c (nCi)

Cs-134 0.9 1.8

Cs-137 0.9 2.1

Co-57 2.1 4.8

Co-60 0.78 1.8

1985–1998 Tl-202 0.5 0.9

C-11/N-13f 0.3 0.5

Eu-152 2.2 3.3

Co-58 0.5 0.9

Co-56 0.5 0.9

Hg-197 3.1 4.6

Hg-195 2.5 3.7

Hg-195m 1.8 3.2

Hg-197m 3.8 6

Hg-203 0.8 1.2

Hg-193m 0.7 1.5

Cs-134 0.5 1.1

Os-185 0.6 1.1

V-45 0.5 0.8

Be-7 3.4 8.7

Sc-46 0.5 0.9

Mn-54 0.5 0.9

Cs-137 0.6 1.1

Co-60 0.5 0.8

Br-77 1.7 3.4

Sb-124 0.4 0.8

Ce-141g 2.2 4.4

Ce-144g 12.1 24.2

Cr-51 6.4 12.8

Co-57 1.4 2.8

Cu-67 1.5 3

Fe-59 1.2 2.4

Se-75 1.1 2.2

Se-73 0.4 0.8

Na-22 0.4 0.8

Zn-65 0.8 1.6

1999–presentg 511 keV 1.15 2.3

Be-7 4.2 8.4

Ce-141g 1.35 2.7

Ce-144g 6.5 13

Co-56 0.55 1.1

Co-58 0.5 1

Co-60 0.45 0.9

Cr-51 4.45 8.9

Cs-134 0.5 1

Cs-137 0.6 1.2

1999–presentg 

(continued)

Cu-67 1.5 3

Eu-152 1.3 2.6

Hg-203 0.6 1.2

Mn-54 0.4 0.8

Na-22 0.45 0.9

Os-185 0.45 0.9

Ra-226 20.5 41

Sb-124 0.45 0.9
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 49 of 150

Period Nuclide LC (nCi) MDAb,c (nCi)

Sc-46 0.7 1.4

Se-75 0.85 1.7

Tl-202 0.5 1

U-235 1.35 2.7

V-48 0.4 0.8

Zn-65 1.1 2.2

Zr-95 0.8 1.6

a. Listing of an MDA for a radionuclide does not necessarily mean that that radionuclide was 

frequently encountered. The MDAs listed in the individual’s results for a given count should 

be used if available.

b. Based on 95% confidence of detection.

c. MDA = LC × 2 unless otherwise specified

d. The HUMCO I and II systems were designed for screening subjects. Subjects found to have 

contamination levels above background were referred to the 8- by 4–in. NaI detector, which 

had the same sensitivities with an extended count time.

e. By bremsstrahlung.

f. Based on 511 keV. C-11 is a positron emitter with no photons. However, the 511-keV peak 

should always be present due to positron annihilation. The 511-keV peak can have 

interference contributions from other sources including pair production interactions from 

nuclides with photon energies greater than 1,022 keV. 

g. Lower sensitivities might be available using the lung counter for certain nuclides if lung 

counting is appropriate to the dose reconstruction.

5.3.1.2 Cesium-137 and Other Intakes from Fallout

Many workers in the early days of whole-body counting had detectable activities of 137Cs. Most of this 

was attributed to fallout. Some workers had even higher levels of 137Cs from consumption of wild 

game. A DL used to establish the difference between occupational and nonoccupational sources of 

137Cs and other fallout radionuclide intakes has not been discovered in the records. In lieu of other 

information, the following guidance can be used: 

• The 137Cs intake should be considered occupational if the same whole-body count detected 

other fission or activation products. It should also be considered occupational if a fission or 

activation product or radiostrontium urinalysis showed detectable activity and the sample was 

obtained in a reasonable time before or after the whole-body count or within the period 

between the previous and next whole-body counts. The reasonable time is based on the 

biological retention pattern of the radionuclide in the body.

• All other fission or activation products that are identified in the whole-body or lung count 

should be considered occupational unless specifically stated on the White Card and the 

reasons for invalidating the results are acceptable to dose reconstruction practices.

• If an investigation occurred and the record clearly shows that the intake was due to a 

nonoccupational source, the 137Cs can be disregarded. The results of the investigation would 

be noted on the White Card.

• National Council on Radiation Protection and Measurements (NCRP) Report 94 provides 

mean body burdens of 137Cs for the United States for the years most likely to produce 

interference with occupational whole-body count results (NCRP 1988). Those values are 

listed in Table 5-19. If no other fission or activation products are linked to the intake and the 

137Cs result is less than the values in Table 5-19, the 137Cs result can be assumed to be due to 

fallout.
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 50 of 150

Table 5-19. Mean body burdens of 137Cs from 

fallout in the United States (nCi).

a

Year Body burden Year Body burden

1953 0.27 1966 9.7

1954 1.1 1967 5.6

1955 2.2 1968 3.5

1956 4.3 1969 2.7

1957 5.1 1970 2.7

1958 6.5 1971 2.7

1959 8.1 1972 2.7

1960 6.8 1973 2.7

1961 4.6 1974 1.6

1962 6 1975 1.1

1963 11 1976 1.6

1964 19 1977 1.1

1965 16

a. NCRP (1988).

5.3.2

Lung burdens of 239Pu, 238Pu, and 241Am were monitored using the phoswich lung detectors beginning 

in 1970. The 59.5-keV gamma line of 241Am is used to determine the 241Am burden (50- to 70-keV 

region). If the isotopic ratio for a given intake is known, the 239Pu and 238Pu can be determined from 

the 241Am. Otherwise, the plutonium is determined from the uranium low-energy X-ray (L X-ray) 

region. When 241Am, 239Pu, and 238Pu are present, corrections for the contribution of the neptunium 

L X-rays, from the decay of 241Am, to the 14- to 25-keV 239Pu and 238Pu regions must be considered. 

The phoswich detectors were eventually replaced by arrays of HPGe detectors, which greatly 

improved the energy resolution. Improved energy resolution permitted the system to distinguish 

between gamma and X-ray lines that are closer together. However, because the uranium L X-ray 

energies for the decay of 239Pu and 238Pu are the same, there is no way to differentiate between these 

two isotopes in an actual measurement. Isotopic information about the exposure is used to determine 

the appropriate calibration factor.

Lung Burdens

Efficiency, and therefore sensitivity level, varies for every count due to the effects of chest wall 

thickness on the attenuation of the 17-keV X-ray and the 59.5-keV gamma ray. Therefore, the MDA 

that is listed with the count should be used when available. The MDA and LC values in Table 5-20 are 

nominal and based on the calibration chest wall thickness of 2.5 cm. These can be used for 

correlation with projected bioassay results. Chest wall thickness for the individual can typically be 

found on the White Card that is associated with the in vivo counting record. Chest wall thickness is 

estimated by weight:height ratios for routine counting and by ultrasound for special or positive counts. 

For lung (chest) counts, increases in chest wall thickness can increase the MDA for the individual 

count. University of California (1977b) suggests that MDAs should be increased 50% for large 

individuals.

Table 5-20. LC and MDA values for lung counting (nCi). 

 The dose reconstructor should use best judgment in determining the applicability of the 

listed MDAs for bounding missed dose projections. The MDA and LC values for lung counting are 

summarized in Attachment A, Table A-26.

Period Radionuclide LC MDAa

1970–1973 (2000 s) (Hempelmann, Richmond, and 

Voelz 1973)

Pu-239 3.5 7

Extended count time

1977b

Am-241 0.3

Pu-238 10

Pu-239 21
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 51 of 150

Period Radionuclide LC MDAa

1980–1984c

(Ennis 2003)d 

(1970–1979)e

Am-241 0.155 0.31

Pu-238 11 22

Pu-239 24 48

1984 

(Vasilik et al. 1984)f

Am-241 0.16 0.32

Pu-238 14 28

Pu-239 30 60

1998g–presenth

(Ennis 2003)

Am-241 0.1 0.2

Am-243 0.1 0.2

Pu-238 10 20

Pu-239 31 62

Th-234 0.85 1.7

U-235 0.1 0.2

Np-237 0.2 0.4

Np-239 0.1 0.2

1998–present

Fission/activation products (Ennis 2003)

511 keV 0.1 0.2

Be-7 0.35 0.7

Ce-141 0.1 0.2

Ce-144 0.25 0.5

Co-56 0.1 0.2

Co-58 0.05 0.1

Co-60 0.1 0.2

Cr-51 0.35 0.7

Cs-134 0.05 0.1

Cs-137 0.1 0.2

Cu-67 0.1 0.2

Eu-152 0.1 0.2

Hg-203 0.05 0.1

Mn-54 0.1 0.2

Na-22 0.1 0.2

Nd-147 0.1 0.2

Os-185 0.05 0.1

Ra-226 0.9 1.8

Sb-124 0.05 0.1

Sc-46 0.1 0.2

Se-75 0.05 0.1

Tl-202 0.05 0.1

a. Assume chest wall thickness of 2.3 cm.

b. As listed in 1977 quarterly progress report based on a 60-minute count time and a person of average build for 

a UPPU Club member (University of California 1977b).

c. Might be applicable to the startup of the phoswich system in 1970; no other information available. For a 

2,000-s count time.

d. Assume chest wall thickness of 2.5 cm.

e. There is a reasonable correlation between the expected MDA for a 15- to 20-minute count time, typical of 

standard in vivo counting beginning in 1970, and the 60-minute count MDAs listed in 1977.

f. MDA = LC × 2; recounts were performed if the results were greater than the LC

g. Lung counter has 10- to 300-keV and 80- to 3,000-keV ranges, so a lower sensitivity for certain fission and 

activation products can be obtained.

h. The phoswich detectors were replaced by an array of six planar HPGe detectors around 1998 (Ennis 2003).

MDAs are typically reported for 239Pu and 241Am for lung counts. However, 241Am is not always 

detectable even if a plutonium lung burden exists. Much of the plutonium at LANL, especially in the 

early years, was fresh [3]. Plutonium burdens from the 26 individuals from the original Manhattan 

Project have been calculated by LANL to be 6 to 80 nCi since before 1946. These individuals have 

been followed for more than 50 years. Twenty-one of them received their first lung counts in 1971. At 

that time the 239Pu MDA was 7 nCi for a 2,000-second count. Positive counts were obtained for 14 of 
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 52 of 150

the 21 measured persons. The lung burdens ranged from 3 to 10 μCi 239Pu (Hempelmann, 

Richmond, and Voelz 1973). None of these individuals has had 241Am activity above background

using in vivo techniques (Voelz et al. 1979; Voelz, Grier, and Hempelmann 1985) except one 

individual with a positive lung count for 241Am 37 years later. This individual was suspected of an 

additional intake of plutonium that contained 241Am in 1957.

Based on the above discussion, ingrowth of 241Am from early intakes (1944 to 1945) should be 

expected to be negligible. Intakes in later years might result in 241Am ingrowth or might have 

contained 241Am in the original intake.

Follow-up lung counts for on members of the UPPU Club were typically longer (60 minutes rather than 

15 or 20) than those for routine or investigational lung counting. Therefore, the MDAs for these 

counts should be expected to be between 50% and 70% of the standard MDAs listed. (The rule of 

thumb is to quadruple the count time and halve the MDA.)

Results in bioassay records should be assumed to be in nanocuries unless otherwise stated. Results 

listed as NULL indicate NDA (not MDA). Results less than the LC are marked as NDA in the 

database. The actual DL values and counting errors might be available on the report of the individual 

but are not available in the database until 2003.

All individuals who receive lung counts are monitored for 239Pu and 241Am. In 2004, 235U and 234Th (as 

238U) were added to the routine in vivo analysis library.

5.3.3

In August 1959, the H-6 Group acquired a probe to be used to monitor wounds that were 

contaminated with plutonium. This probe was capable of detecting soft plutonium X-rays. The 

sensitivity of this probe was 1 × 10-9 Ci of plutonium when it was unshielded. This was equivalent to 

detection of one-tenth of a permissible body burden of embedded plutonium in tissue to a depth of 

1 cm (Shipman 1959). In 1977, a new NaI detector (12 by 2 mm) was being evaluated. This 

produced an MDA of 0.07 nCi based on WGPu (University of California 1977c).

Wound Monitoring

Wound counting was used primarily as a tool for surgeons to locate plutonium in the wound, not as 

results for calculation of internal dose. Wound monitoring continues to be performed. In most cases, 

intake and dose are not assessed directly from the wound count but rather from urine bioassay data. 

When necessary, wound results can be used to estimate doses using the guidance in ORAUT-OTIB0022, Guidance on Wound Modeling for Internal Dose Reconstruction (ORAUT 2005b). Follow-up 

studies of wounds found that plutonium does not readily migrate from the wound site to uptake in the 

majority of the incidents (Voelz and Lawrence 1991). No other information on instrumentation or 

sensitivities is available.

5.4 INTERFERENCES AND UNCERTAINTIES

5.4.1

In the early years of the bioassay program, analysis techniques and protocols did not use chemical 

blanks and recoveries in the manner that became standard protocol in later years. When the 

databases were being constructed, beginning in the late 1960s, an attempt was made to incorporate 

parameters in relation to current analysis procedures (Glover 2006). Therefore, the results initially 

reported to the worker might not be the same as the results that finally appear in the database that 

was submitted to NIOSH. A project to validate these results is described in ORAUT-OTIB-0063, Los 

Alamos National Laboratory Bioassay Repository Database (ORAUT 2009a). If there are 

Discrepancies in Results
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 53 of 150

discrepancies between results in the current database and the results supplied by the worker, the 

dose reconstructor should refer to the results of the validation project to resolve any discrepancies.

5.4.2

Until the Health Pass Ward was established in 1945, the potential existed for sample contamination. 

It is likely that a contaminated sample will appear as an obvious outlier in the dataset for a worker. 

The use of a result from a contaminated sample could result in an overestimate of the worker’s dose, 

but the sample result should be considered real if other data do not demonstrate it to be a false 

positive result. Some variability is expected in any set of results. See Section A.9, Attachment A, for 

an example of a valid series of bioassay results for an individual whose autopsy results verified 

plutonium intakes. The Health Pass Ward was eliminated in 1952 because of expense. However, 

protocols to allow the samples to be collected in a contamination-free environment were established.

Contamination of Samples

For in vivo measurements, contamination can occur as external to the body or, in the case of chest 

counting, as external to the lung. If a follow-up in vivo count shows a dramatic decrease in activity or 

no detectable activity on the same day or within a few days, external contamination can be suspected. 

Radon progeny and medical diagnostic or therapeutic procedures that involve radionuclides can 

cause interference to in vivo measurements, especially for NaI detectors. However, unless the count 

was invalidated or noted as being influenced by such interferences, the results should be used as 

recorded.

5.4.3

5.4.3.1 In Vivo Counting

Uncertainties

Uncertainties for bioassay measurements might be included with the results for excreta or in vivo

measurements. For in vivo results, uncertainties are not reported in the database until 2003, but 

uncertainties might be listed on individual reports before 2003. The listed uncertainties are typically 

reported as 1 standard deviation.

Uncertainties that are associated with chest counting are reduced by the use of different calibrations 

for different chest wall thicknesses and the use of ultrasound to measure chest wall thickness. A 

1-sigma uncertainty of approximately 20% for americium and uranium values in chest counting, not 

including correction for interferences from bone and liver, is assumed. Uncertainties would be much 

higher for an individual with activity in the bone or liver. The uncertainty in lung activity estimates that

are affected by contributions from activity in the liver and skeleton would probably range from 100% or 

more for levels near or below the MDA to 50% or more for activity above the MDA. The uncertainty in 

the estimate of chest thickness using the height:weight correction was at least 50% (University of 

California 1977b). The mathematical correction was made for routine counts. Special counts and 

counts with positive results were typically corrected using ultrasound chest wall thickness 

measurements.

5.4.3.2 In Vitro Measurements

Uncertainties for normal distributions are often listed in the Bioassay Repository database for 

samples. The measurement uncertainty is typically listed at 1 standard deviation with the results

(Lewis 2006a). When results are listed as negative, positive, or zero, use of a normal distribution is 

recommended because a lognormal distribution would not be applicable to negative results. 

However, a study of the variations in plutonium urinary data (Moss et al. 1969) determined that the 

LANL employee results exhibited as a lognormal distribution with a geometric standard deviation 

(GSD) of 1.9 (from the ratio of the 84th percentile to the 50th percentile). 
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 54 of 150

5.5 UNMONITORED INTAKES

Exposure circumstances discussed throughout this TBD indicate that in every case for Los Alamos 

National Laboratory employees, the potential for unmonitored potential intakes must be evaluated. In 

some cases, unmonitored exposures must be inferred from a review of the claim information. The 

most useful sources of information for this review (beyond internal dose monitoring records) are listed 

below.

5.5.1.1 Energy Employee Interview

The interview record offers the energy employee the opportunity to choose from a list of radionuclides 

to which he or she might have been exposed, and offers the opportunity to include others that are not 

listed. In addition, the employee is asked about potential exposures and to provide details of 

exposures in relation to job assignments or incidents.

5.5.1.2 Survivor or Less-Detailed Interview

The quality of information about exposures is much lower when the interviewee is a survivor rather 

than the energy employee. Even many energy employees did not have a detailed knowledge of what 

they might have been exposed to, or if they had, they might not remember. Some employees check 

every radionuclide in the list; this is not typically a credible set of exposures because not all of the 

radionuclides in the interview list were present at LANL.

5.5.1.3 Energy Employee, Science Job Category Interview

Often LANL interviewees have excellent recollections of their activities. In dose reconstructions that 

have been completed to date, this has been shown to be particularly true of researchers and 

scientists, who typically listed the radionuclides to which they were most exposed. What is more, the 

individuals most likely to be exposed to LANL’s less common radionuclides are researchers and 

scientists because routine operations for processing did not generally occur on these materials.

5.5.1.4 Initial Case File

Another excellent resource for identifying work assignments that could result in unmonitored potential 

exposures is the initial case file supplied by the DOL. This file often has records that might list 

detailed job assignments. This information can be used to define a potential unmonitored exposure.

Once the potential for an unmonitored intake is identified, a variety of techniques can be applied to 

estimate doses from unmonitored intakes as described in this section.

Assignment of coworker dose is done in accordance with the sections below, which discuss specific 

radionuclides, and with ORAUT-OTIB-0060, Internal Dose Reconstruction (ORAUT 2007c). 

Evaluation of the values in Attachments B and F should be performed before assignment of dose from 

coworker intakes to see if any correspond with the potential intake.

5.5.2

Monitoring for primary radionuclides has been performed in most years at LANL, although bioassay 

results for americium are sparse before the common use of in vivo measurements from the 1970s. As 

described in Section 5.1.5, some work groups or individuals might not have been monitored or might

have been incompletely monitored, even for primary radionuclides. For unmonitored intakes of 

primary radionuclides that are implied in case records, except americium, dose might be assigned for 

Unmonitored Intakes of Primary Radionuclides
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 55 of 150

using the coworker dose intakes in Internal Dosimetry Coworker Data for Los Alamos National 

Laboratory (ORAUT 2009b). Reconstruction of internal dose from intakes of americium has been 

determined to be infeasible for periods before January 1, 1976 (NIOSH 2007b); americium dose 

assignment from this date forward is discussed in the next section. Intakes are scaled based on job 

category and duration of potential intake using the methodology in Attachment H of this TBD. 

From 1943 through 1946 there was a significant potential for unmonitored intakes of plutonium 

(nominally 3% fresh mixture), 210Po in designated areas, and uranium (nominally either DU or EU 

even though NU was used extensively in conventional weapons testing). 

5.5.2.1 Plutonium

In accordance with the designation of the class of employees added to the SEC, reconstruction of 

internal dose from intakes of plutonium has been determined to be infeasible for periods prior to 1944

(NIOSH 2007b). Bioassay for plutonium was not available before late 1944 and was not completely 

established until 1945. If a worker had a potential for plutonium exposure from 1943 through 1946, 

with no bioassay during that period but plutonium bioassay during later years, dose reconstructors 

may, in some cases, use the later plutonium bioassay with assumptions that are favorable to 

claimants to bound intakes (Schulte and Meyer 1957). However, if the worker did not participate in a 

plutonium bioassay program after this period, coworker dose intakes from ORAUT-OTIB-0062 

(ORAUT 2009b) should be assigned from January 1, 1944, onward for plutonium-239 mixtures. .

5.5.2.2 Polonium

Polonium was used in nuclear weapons initiators, and the primary manufacture of the parts took place 

at the Monsanto Chemical Company. In February 1945, the schedule for polonium delivery from 

Monsanto to TA-1 was increased from a few curies to 100 Ci/mo by June and 500 Ci/mo by December. 

At TA-1, polonium and radium were handled in H and Gamma Buildings. H Building was used for 

preparation of neutron and alpha sources for initiators and isotopic experiments. Workers who were 

involved in these operations were part of Group CM-15. Operations in H Building were carried out 

between 1943 and July 1945 when operations were moved to the DP Site. Polonium-210 was 

processed through various operations that included (1) solution chemistry, (2) electrodeposition, (3) 

high-vacuum distillation, (4) metal plating, and (5) counting and assay of polonium.

Intakes of polonium might be Type F or M, and unless bioassay data are precisely fit, the favorable to 

claimant absorption type should be assumed. Intakes of polonium do not exhibit a long-term 

excretion in the same manner as plutonium intakes; therefore, dose assessment for early intakes

using bioassay data from a later time is not possible.

For polonium, existing bioassay results were used to estimate coworker dose intakes for 1944 to 

1956. Before 1944, the reconstruction of dose from intakes of polonium was determined to be 

infeasible (NIOSH 2007b).

5.5.2.3 Uranium

Coworker dose intakes for uranium are available beginning in 1947. For exposure prior to 1947, the 

uranium intakes assigned for atomic weapons employers conducting uranium machining activities 

should be assigned to those workers who may have been involved with uranium machining (Battelle 

2006) as given in Table 5-21. These intake rates are applicable to machining work involving natural 

uranium. Potential internal exposure due to machining activities involved highly enriched uranium 

cannot be reconstructed. 
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 56 of 150

Table 5-21. Hypothetical chronic intakes for uranium, 1943 to 1946. 

Job Category Intake, pCi/d

Machinist 19654

General Laborer 9827

Supervisor 4914

Clerical 491

5.5.2.4 Tritium

Coworker dose intakes have been calculated for LANL as described in ORAUT (2009b) and cover 

1950 and onward. Reconstruction of tritium doses has been determined to be infeasible before the 

year 1950 (NIOSH 2007b).

5.5.3

In evaluation of the proposed addition of a class of LANL employees to the SEC (with employment 

from 1943 to 1975), a number of radionuclides were identified for which it was not feasible to 

reconstruct doses (NIOSH 2007b). These are listed in Table 5-22. 

Unmonitored Intakes of LANL Radionuclides other than Primary Radionuclides

Table 5-22. Radionuclides for which 

reconstruction is not feasible. 

Radionuclide Dates of infeasibility

3

H 1943–1949

227Ac 1943–1975

244Cm 1943–1975

241Am 1943–1975

237Np 1943–1975

231Pa 1943–1975

230,232Th 1943–1975

MFPs/MAPs 1943–1975

In the post-1975 time frame through the present, bioassay capabilities for these radionuclides have 

existed (Hoover 2008). Since the mid-1970s, LANL has maintained consistency in its approach to 

internal dose monitoring. As described by the LANL program (Hoover 2008): 

...internal dose monitoring is prescribed based on the likelihood of exposure. Routine 

internal dosimetry programs...are established and implemented based on worker 

locations and activities, tailored to the types, quantities, and potential for intake of 

radioactive materials. Special internal dosimetry measurements are made under 

unique circumstances—typically radiological events—where additional or unique 

measurements are warranted, including when unusual results are obtained.”

As listed in the reference, LANL maintains the following bioassay capabilities: 

• Actinium-227

•

. LANL has the ability to detect 227Ac with target in vivo measurements. 

Targeted dosimetry was performed in the past, but no intakes were detected.

Americium-241. A large number of bioassay results exist for this radionuclide in the post-1975 

period.
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 57 of 150

• Neptunium-237

•

: A routine threshold for internal dose monitoring exists in LANL TBDs. A 

contract exists to process in vitro samples, and the ready ability to perform in vivo 

measurements exists.

Protactinium-231

•

. LANL can detect 231Pa with in vivo measurements if warranted.

Curium-244

•

. LANL maintains in vitro capability with targeted bioassay and analysis of 

samples at ORNL.

Thorium-230 and -232

•

. LANL maintains an accredited in vitro bioassay program under the 

DOE Laboratory Accreditation Program using a contract laboratory for sample analysis. In 

vivo bioassay might be conducted on the site, and 232Th has been detected using in vivo

analysis.

Strontium-90

Based on the foregoing, the potential for an unmonitored intake of the listed radionuclides is limited; 

however, it is necessary to demonstrate that maximum doses can be assigned when presumptive 

intakes, beyond those that might be characterized by bioassay records, are implied by case-specific 

information. When doses cannot be assigned from information in Attachments B and F or 

overestimated using ORAUT (2005a), and for the post-1975 era, it is necessary to assign doses using 

coworker dose intakes in ORAUT-OTIB-0062, Internal Coworker Dose Intakes for the Los Alamos 

National Laboratory (ORAUT 2009b).

: LANL is able to detect this radionuclide using in vitro analysis. Targeted 

bioassay has been conducted in the past.

A short description of the work in relation to each of the listed radionuclides follows. Most were 

handled as part of basic research, and the number of exposed individuals would have been small.

5.5.3.1 Actinium-227

Several buildings in TA-21 were used for work with 227Ac that was used as a substitute for the 210Po 

used in weapons initiators. Weapons that used initiators are likely to have been retired by about 

1963, and it is not known whether actinium was used as part of a weapon that was actually deployed. 

The actinium production program was cancelled in 1955 (DOE 1993), and processing activities at 

LANL would have ended by 1955 or 1956.

Actinium-227 bioassay results exist for LANL workers only for 1954, but have not to date been 

available to the dose reconstructor. A memorandum from the U.S. Atomic Energy Commission (AEC) 

that described LANL releases indicates that 2.5 Ci of 227Ac had been disposed in covered absorption 

beds, with a decay-corrected inventory of 1.4 Ci as of November 11, 1973 (Wingfield 1974). The 

originally disposed amount had decayed to approximately 56% of its original value, which implies a 

residence time of approximately 18 years in the burial ground. This would mean that the material was 

disposed of in 1955 or 1956. 

A filter building was constructed in the late 1940s to process exhaust air from Buildings 146 and 152 

in TA-21 (Harper and Garde 1981). The likely primary source of any intake of actinium in the period 

after 1975 was the decontamination and decommissioning of this facility. Health physics techniques 

that were applied to the decommissioning and decontamination operation likely controlled the 

potential intakes of workers, and potential intakes to individuals outside this group would be negligible. 

Table 5-23 lists recommendations.

Table 5-23. Actinium-227 dose assignment recommendations. 
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 58 of 150

Period of potential exposure 1943-1975

Locations of potential exposures DP-21-146, DP-21-152

Dose estimation Dose reconstruction infeasible

Period of potential exposure March-June 1978

Locations of potential exposures DP-21, Building 153

Job titles of personnel potentially exposed Decommissioning and decontamination workers; see 

reference for relative time spent on project. 

Dose estimation techniques (in order of 

preference) 

1. Base on measured air concentrations from Harper and 

Garde (1981). 

2. Assign Ac-227 dose based on coworker intakes for Pu-239.

5.5.3.2 Americium

Although reconstruction of dose from americium before 1976 has been determined to be infeasible 

(NIOSH 2007b), dose reconstruction for the period after 1975 is straightforward due to the availability 

of personnel monitoring results, which are primarily from in vivo monitoring. Table 5-24 lists 

recommendations. 

Table 5-24. Americium dose assignment recommendations. 

Period of potential exposure Before January 1, 1976

Dose estimation Dose reconstruction infeasible

Period of potential exposure January 1, 1976, and after

Locations of potential exposures Various

Job titles of personnel potentially exposed Actinide researchers, process operators, decontamination 

workers, janitors, others. 

Dose estimation technique 1. Base on in vivo measurement results.

2. Assign Am-241 dose based on coworker intakes for Pu-239.

5.5.3.3 Curium-244

The effort to isolate actinides, along with investigation into their chemistry, was conducted partly at 

LANL, and it appears that use of this radionuclide related primarily to actinide research (Penneman 

and Keenan 1960). Research was conducted in the Medical Laboratory in TA-1 and in the CMR-4 

laboratory at DP West in TA-21. A number of papers were published on the radiochemistry of 

americium and curium that span a period from 1958 to 1967 (these were identified in searches of the 

Office of Science and Technical Information electronic archive using keywords americium and 

curium). A 1974 AEC memorandum that identified airborne releases in the period before 1967 lists a 

one-time release of 244Cm from Building 42 in TA-1 (the medical laboratory or ML Building) over the 

period from 1944 to 1956 (Wingfield 1974). Curium bioassay results are available for 1955 in a 

bioassay logbook (LANL 1954–1957) but are not known to be in claimant records.

In 2003, a worker’s DOE badge was found to be contaminated with 244Cm; however, rather than 

suggesting a generalized exposure hazard, the material was traced back to legacy contamination of 

objects that had been disposed as part of a remediation activity for the hot cell area of TA-48-1 (the 

Radiochemistry Laboratory). The material had been stored for tens of years and a sample from the 

box was labeled in the 1970s (DOE 2004). In this case, in vivo and in vitro bioassay using the RAS 

technique were performed; the individual’s dose was estimated to be approximately 0.01 rem (Little 

et. al. 2003c). Table 5-25 lists recommendations.
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 59 of 150

Table 5-25. Curium-244 dose assignment recommendations. 

Period of potential exposure Before January 1, 1976

Dose estimation Dose reconstruction infeasible

Period of potential exposure January 1, 1976, and after

Locations of potential exposures CMR Building, Radiochemistry Laboratory

Job titles of personnel potentially exposed Researchers, decommissioning and decontamination workers. 

Dose estimation techniques (in order of 

preference) 

1. Base on bioassay results.

2. Assign Cm-244 dose based on coworker intakes for Pu238.

5.5.3.4 Neptunium-237

Limited information was found about neptunium at LANL, but It appears that 237Np activities have 

primarily been associated with the Nuclear Fuels Group. The 237Np metal for the bare criticality 

experiment in 2002 (Roark 2003) was prepared at the CMR Building (University of California 1996). 

At that time, LANL maintained a routine dosimetry threshold for 237Np in internal dosimetry TBDs 

(Hoover 2008); however, no workers at that time were on routine bioassay programs. This is likely 

due to the isolation that is afforded by the hot cells facilities in Wing 9 of the CMR Building. These 

facilities had been available at LANL since their dedication in 1961 (University of California 1996). 

Table 5-26 lists recommendations.

Table 5-26. Neptunium-237 dose assignment recommendations. 

Period of potential exposure Before January 1, 1976

Dose estimation Dose reconstruction infeasible

Period of potential exposure January 1, 1976, and after

Locations of potential exposures CMR, Wing 9.

Job titles of personnel potentially exposed Actinide researchers, nuclear fuels technicians, process 

operators, decontamination workers, janitors, others. 

Dose estimation techniques (in order of 

preference) 

1. Base on bioassay results.

2. Assign Np-237 dose based on coworker intakes for Pu-239. 

5.5.3.5 Protactinium-231

Papers on the chemistry of protactinium were published from 1959 to 1979 (Kirby 1959; Asprey and 

Penneman 1964; Smith, Spirlet, and Muller 1979). The foundation work is documented in 

Radiochemistry of Protactinium, which is based on work performed at the Mound site (Kirby 1959). 

Another Mound document (Eppley and Valleé 1979) shows that LANL had requested 231Pa in the 

amounts of 0.1, 30.1, and 11.0 g in fiscal years 1979, 1980, and 1981, respectively, but the pilot plant 

producing the material was shut down in 1979 (DOE 1993). Total production for the plant was less 

than 0.9 g (DOE 1993), so the amounts requested for years 1980 and 1981 could not have been 

delivered . No production-scale operations were identified at LANL, and there is no mention of the 

material in airborne radioactivity records or in the waste inventories in the 1974 AEC memorandum 

(Wingfield 1974). In vitro bioassay methodologies are described for 231Pa in Attachment C of this 

TBD, and LANL has maintained the ability to detect this radionuclide since the mid-1970s (Hoover 

2008). Table 5-27 lists recommendations.
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 60 of 150

Table 5-27. Protactinium-231 dose assignment recommendations. 

Period of potential exposure Before January 1, 1976

Dose estimation Dose reconstruction infeasible

Period of potential exposure January 1, 1976, and after

Locations of potential exposures Various

Job titles of personnel potentially exposed Actinide researchers, decontamination workers, janitors. 

Dose estimation techniques (in order of 

preference) 

1. Base on bioassay results.

2. Assign Pa-231 dose based on coworker intakes for Pu-239. 

5.5.3.6 Thorium

Two distinct source terms are associated with 230Th and 232Th at LANL. 

5.5.3.7 Thorium-230

In the 1950s an effort was underway within the AEC complex to build a production capacity for 230Th, 

which was also called ionium, and a projected use of the material was for radioisotope heat sources. 

This work seems to be associated with a number of sites as part of a general effort sponsored by the 

AEC. No records were found among the available literature that indicated the presence of isolated 

material other than a number of bioassays for 230Th in 1958 found in Los Alamos National Laboratory 

bioassay records. In December 1956, the Mound plant shipped 49 g of 230Th to the University of 

California Radiation Laboratory in Berkeley (DOE 1993). The material does not compose a part of the 

airborne release or waste disposal inventories that accumulated through 1973 in the 1974 AEC

memorandum (Wingfield 1974). Table 5-28 lists recommendations.

Table 5-28. Thorium-230 dose assignment recommendations. 

Period of potential exposure Before January 1, 1976

Dose estimation Dose reconstruction infeasible

Period of potential exposure January 1, 1976, and after

Locations of potential exposures Exposure unlikely

Dose estimation techniques (in order of 

preference) 

1. Base on bioassay results.

2. Assign Th-230 dose based on coworker dose intakes for 

uranium. 

5.5.3.8 Thorium-232

It appears that there might have been two periods when 232Th work was conducted. The first period 

might be associated with a potential airborne exposure from the Sigma complex over the period from 

1944 to 1963, and 232Th was one of the identified radionuclides (Wingfield 1974). Casting, machining, 

and powder metallurgy were conducted at the Sigma complex, which housed the Material Science 

and Technology Division. This facility was built in the 1950s and 1960s (LANL 2007). The material 

was in the form of ingots and oxides in the building designated as a Thorium Storage Building 

(Building 159). Building 66 in TA-3 was a warehouse where thorium was stored, and air emissions 

results from several years indicate airborne releases of 232Th from this location (Wingfield 1974).

A 1958 reference describes safety measures for uranium and thorium (Stout 1958). The procedures 

in the reference apply to both elements due to the similarity of the hazards (radioactivity, toxicity, 

explosion, and fire). The comment is made in this publication that this manual was directing a 

changed policy toward the handling of thorium, and that it was to be considered in the future an active 

alpha emitter (Stout 1958). The implication is that 232Th, with its very long half-life, had been 

considered essentially stable. It is true that the activity is low for this reason, but the ingrowth of its 
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 61 of 150

alpha-emitting progeny occurs within a short period based on their short half-lives. Because the 

material had not been assumed to be a hazard before this statement, the material is assumed to 

consist of 232Th rather than the more active 230Th, which with its higher specific activity would have 

required controls with other active alpha emitters.

Later work with 232Th appears to have taken place in the 1980s (ORAUT 2004). During this period, 

much more was known about the potential hazards of thorium, and LANL has used in vivo and in vitro

techniques to monitor for intakes of the material. Table 5-29 lists recommendations.

Table 5-29. Thorium-232 dose assignment recommendations. 

Period of potential exposure Before January 1, 1976

Dose estimation Dose reconstruction infeasible

Period of potential exposure January 1, 1976, and after

Locations of potential exposures TA-3-66

Job titles of personnel potentially exposed Materials researchers and processing workers. 

Dose estimation techniques (in order of 

preference) 

1. Base on bioassay results.

2. Assign Th-232 dose based on coworker intakes for 

uranium.

5.5.3.9 Mixed Fission and Activation Products

MFPs are typically associated with reactors or with facilities where irradiated reactor fuel is processed. 

Mixed activation products (MAPs) are the result of neutron irradiation in a reactor core or an 

accelerator. Before the advent of whole-body counting, radiation protection policy for MFPs was often 

to collect bioassay samples for gross beta and assign the dose to the most limiting radionuclide. For 

this reason, bioassay results for a spectrum of radionuclides from this source term are not expected, 

and gross beta or isotopic samples for 90Sr and 137Cs are typical. The in vivo monitoring program is 

capable of detecting a large range of fission and activation products. Table 5-30 lists 

recommendations.

Table 5-30. MFP and MAP dose assignment recommendations. 

Period of potential exposure Before January 1, 1976

Dose estimation Dose reconstruction infeasible

Period of potential exposure January 1, 1976, and after

Locations of potential exposures Reactor facilities, accelerator facilities

Job titles of personnel potentially exposed Non-administrative workers assigned to these areas. 

Dose estimation techniques (in order of 

preference)

1. Records indicate an extensive program of in vivo 

bioassay for various fission and activation products; claim 

records likely contain bioassay records to reconstruct 

these doses.

2. Assign dose from radionuclide of interest using the 

coworker intakes for Cs-137. 

5.6 ATTRIBUTIONS AND ANNOTATIONS

Where appropriate in this document, bracketed callouts have been inserted to indicate information, 

conclusions, and recommendations provided to assist in the process of worker dose reconstruction. 

These callouts are listed here in the Attributions and Annotations section, with information to identify 

the source and justification for each associated item. Conventional References, which are provided in 

the next section of this document, link data, quotations, and other information to documents available 

for review on the Project’s Site Research Database (SRDB).
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 62 of 150

 [1] William Moss. University of California (retired). Radiochemist. 2007.

Discussions with William Moss indicate that no correction for recovery was included in the 

calculation of the polonium activity. There is no indication that any further recalculations were 

performed on the bioassay results before loading these results in the current bioassay 

database.

[2] Elizabeth Brackett. ORAU Team. Principal Internal Dosimetrist. 2007.

E-mail correspondence indicates that the polonium in urine procedure by Monsanto used a 

recovery factor of 85%, when instead the factor should have been 10%.

[3] James Lawrence. University of California (retired). Health Physicist. 2006.

Discussions with James Lawrence indicate that most of the early work with plutonium was with 

what he described as fresh plutonium.
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 63 of 150

REFERENCES

Adley, F. G., C. M. Berry, E. W. Gilliland, E. C. Hyatt, F. R. Ingram, H. S. Jordan, Jr., S. J. Pearce, 

H. H. Shrenk, and L. Silverman, 1963, Respiratory Protective Devices Manual, Chapter 11,

Special Applications, American Industrial Hygiene Association, Braun and Brunfield, Ann 

Arbor, Michigan. [SRDB Ref ID: 7963]

Asprey, L. B. and R.A. Penneman, 1964, “Protactinium Fluorides, the New Class, MPaF6,” Science, 

volume 145, number 3635, p. 934, August 28. [SRDB Ref ID: 41604]

Bates, B., 1998, Summary of Radiological Incident Reports (RIR) through June 1998, Los Alamos 

National Laboratory, Los Alamos, New Mexico, July. [SRDB Ref ID: 883]

Battelle, 2006, Site Profiles for Atomic Weapons Employers that Worked Uranium and Thorium 

Metals, Battelle- TBD-6000, PNWD-3738, Revision 0, December 13. 

Buckland, C., 1958, 1957 Annual Progress Report, H-1 General Monitoring Section, Los Alamos 

Scientific Laboratory, Los Alamos, New Mexico, January 22. [SRDB Ref ID: 27969]

Buckland, C., 1959, 1958 Annual Progress Report, H-1 General Monitoring Section, Los Alamos 

Scientific Laboratory, Los Alamos, New Mexico, March 10. [SRDB Ref ID: 27989]

Buckland, C., 1961, 1960 Progress Report, H-1 General Monitoring Section, Los Alamos Scientific 

Laboratory, Los Alamos, New Mexico, April 6. [SRDB Ref ID: 27991]

Buddenbaum, J., 2006a, “LANL Lab Notebooks,” e-mail correspondence, ENSR|AECOM, Westford, 

Massachusetts, September 1. [SRDB Ref ID: 27995]

Buddenbaum, J., 2006b, “Conference Call Minutes, Final,” ENSR|AECOM, Westford, Massachusetts, 

February 13. [SRDB Ref ID: 27992]

Burke, J. E., 1946, Report of Plant Planning Committee Meeting, March 8, 1946, Los Alamos 

Scientific Laboratory, Los Alamos, New Mexico, March 9. [SRDB Ref ID: 27976]

Campbell, E. E., M. F. Milligan, W. D. Moss, and H. F. Schulte, 1972, History of the Plutonium 

Bioassay Program at the Los Alamos Scientific Laboratory, 1944-1972, LA-5008, University of 

California, Los Alamos Scientific Laboratory, Los Alamos, New Mexico, October. [SRDB Ref 

ID: 941]

Cember, H., 1983, Introduction to Health Physics, 2nd Edition, Pergamon Press, Elmsford, New York.

Cheng, Y. S., R. A. Guilmette, Y. Zhou, J. Gao, T. LaBone, J. J. Whicker, and M. D. Hoover, 2004, 

“Characterization of Plutonium Aerosol Collected During An Accident,” Health Physics, volume 

87, number 6, pp. 596–605. [SRDB Ref ID: 25247]

Christensen, E. L., R. Garde, and A. M. Valentine, 1975, Demolition of Building 12, An Old Plutonium 

Filter Facility, LA-5755, University of California, Los Alamos Scientific Laboratory, Los Alamos, 

New Mexico, January. [SRDB Ref ID: 27934]

Clark, J. M., 2005, History of LANL’s Bioassay Program from Inception to 1993, LA-UR-05-1942, Los 

Alamos National Laboratory, Los Alamos, New Mexico, April. [SRDB Ref ID: 17157]
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 64 of 150

Cox, W. S., 1945, “Prevention of any Personnel Entering Contaminated Air Areas in DP East and 

West,” memorandum, University of California, Los Alamos Scientific Laboratory, Los Alamos, 

New Mexico, December 4. [SRDB Ref ID: 27962]

Currie, L. A., 1968, “Limits for Qualitative Detection and Quantitative Determination, Application to 

Radiochemistry,” Analytical Chemistry, volume 40, number 3, pp. 586–593. [SRDB Ref ID: 

12372]

DOE (U.S. Department of Energy), 1988, Radiation Protection for Occupational Workers, Order

5480.11, Office of Environment, Safety and Health, Washington, D.C., December 21. [SRDB 

Ref ID: 8073]

DOE (U.S. Department of Energy), 1991, Tiger Team Assessment of the Los Alamos National 

Laboratory, Washington, D.C., November. [SRDB Ref ID: 23620]

DOE (U.S. Department of Energy), 1993, Operable Unit 9, Site Scoping Report: Volume 7 - Waste 

Management, Final Report (Revision 0), Albuquerque Field Office, Albuquerque, New Mexico, 

February. [SRDB Reference ID 11877]

DOE (U.S. Department of Energy), 2004, (Occurrence Report) Curium-244 Detected on Employee’s 

DOE Badge, Badge Transported Off-Site, Improper Shipment, DP-ALO-LA-LANL-ESHSUPT2003-0001, Albuquerque Field Office, Albuquerque, New Mexico, July 6. [SRDB Ref ID: 

57071] 

Dummer, J. E., editor, 1958, Los Alamos Handbook of Radiation Monitoring, LA-1835 (3rd Edition), 

University of California, Los Alamos Scientific Laboratory, Los Alamos, New Mexico, 

November. [SRDB Ref ID: 967]

Dummer, J. E., 1961a, “Iodine Activity from DP West Hot Cells,” memorandum to D. Meyer, University 

of California, Los Alamos Scientific Laboratory, Los Alamos, New Mexico. June 2. [SRDB Ref 

ID: 14523]

Dummer, J. E., 1961b, “Iodine Activity from DP West Hot Cells,” memorandum to D. Meyer, University 

of California, Los Alamos Scientific Laboratory, Los Alamos, New Mexico, June 2. [SRDB Ref 

ID: 14523]

Ennis, M., 2003, Historical Minimum Detectable Activities for LANL’s In Vivo Measurements 

Laboratory, LA-UR-03-7702, Los Alamos National Laboratory, Los Alamos, New Mexico. 

[SRDB Ref ID: 27304]

ENSR (ENSR International), 2002, A Summary of Historical Operations at Los Alamos National 

Laboratory and Associated Off-Site Releases of Radionuclides and Other Toxic Materials, 

Version 2G, Draft, Alameda, California, February 22. [SRDB Ref ID: 27946]

Eppley, R. E. and R. E. Vallee, 1979, Reported Needs for the Radioactive Isotopes Uranium-234, 

Protactinium-231, Thorium-230, and Thorium-229, MLM-ML-77-46-0003, Monsanto Research 

Corporation, Mound Laboratory, Miamisburg, Ohio, June 1. [SRDB Ref ID: 25195]

Foreman, H., W. Moss, and W. Langham, 1960, “Plutonium Accumulation from Long-Term 

Occupational Exposure,” Health Physics, volume 2, pp. 326–333. [SRDB Ref ID: 12887]
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 65 of 150

Gallaher, B. M., and D. E. Efurd, 2002, Plutonium and Uranium from Los Alamos National Laboratory 

in Sediments of the Northern Rio Grande Valley, LA-13974, University of California, Los 

Alamos National Laboratory, Los Alamos, New Mexico, August. [SRDB Ref ID: 12963]

Gautier, M. A., 1983, Manual of Analytical Methods for Radiobioassay, LA-9763-M, University of 

California, Los Alamos National Laboratory, Los Alamos, New Mexico, July. [SRDB Ref ID: 

12890]

Glover., S., 2006, “Data Discrepancy at LANL; emailing LA-345.pdf,” e-mail to J. Buddenbaum, 

National Institute for Occupational Safety and Health, Cincinnati, Ohio, August 30. [SRDB Ref 

ID: 27998]

Harper, J. R., and R. Garde, 1981, The Decommissioning of TA-21-153, A 227Ac Contaminated Old 

Filter Building, LA-9047-MS, Los Alamos National Laboratory, Los Alamos, New Mexico, 

November. [SRDB Ref ID: 929]

Harper, J. R., and R. Garde, 1982, The Decommissioning of a Tritium-Contaminated Laboratory, LAUR-82-2323, Los Alamos National Laboratory, Los Alamos, New Mexico, November. [SRDB 

Ref ID: 70272] 

Healy, J. W., 1970, 1970 Los Alamos Handbook of Radiation Monitoring, LA-4400, University of 

California, Los Alamos Scientific Laboratory, Los Alamos, New Mexico. [SRDB Ref ID: 943]

Hempelmann, L., 1944, Health Report, LAMS-81, Los Alamos, New Mexico, April 14. [SRDB Ref ID: 

881]

Hempelmann, L. H., 1946, History of the Health Group (A-6), (March 1943 to November 1945), 

University of California, Los Alamos Scientific Laboratory, Los Alamos, New Mexico, April 6. 

[SRDB Ref ID: 12894]

Hempelmann, L. H., and W. H. Langham, 1953, Determination of Systematically Deposited Plutonium 

in Laboratory Personnel and a Simple Qualitative Test for Exposure to Airborne Radioactive 

Material, AECU-2633, University of California, Los Alamos Scientific Laboratory, Los Alamos, 

New Mexico. [SRDB Ref ID: 14518]

Hempelmann, L. H., C. R. Richmond, and G. L. Voelz, 1973, A Twenty-Seven Year Study of Selected 

Los Alamos Plutonium Workers, LA-5148-MS, University of California, Los Alamos Scientific 

Laboratory, Los Alamos, New Mexico, January. [SRDB Ref ID: 938]

Hoover, P., 2007, “In Vivo Counting, 2007,” e-mail to J. Buddenbaum, Los Alamos National 

Laboratory, Los Alamos, New Mexico, January 3. [SRDB Ref ID: 28600]

HPS (Health Physics Society), 1996, An American National Standard – Performance Criteria for 

Radiobioassay, ANSI Standard N13.30-1996, American National Standards Institute, New 

York, New York. 

Hughes, S. D., 2000, The Unclosed Circle: Los Alamos and the Human and Environmental Legacy of 

the Atom, 1943-1963, University of New Mexico, Albuquerque, New Mexico, May. [SRDB Ref 

ID: 48438]
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 66 of 150

ICRP (International Commission on Radiological Protection), 1975, Report of the Task Group on 

Reference Man: Anatomical, Physiological and Metabolic Characteristics, Publication 23, 

Pergamon Press, Oxford, England.

ICRP (International Commission on Radiological Protection), 1979, Limits for the Intakes of 

Radionuclides by Workers, Part 1, Publication 30, Part 1, Pergamon Press, Oxford, England.

ICRP (International Commission on Radiological Protection), 1994, Human Respiratory Tract Model 

for Radiological Protection, Publication 66, Pergamon Press, Oxford, England.

ICRP (International Commission on Radiological Protection), 1995, Dose Coefficients for Intakes of 

Radionuclides for Workers, Publication 68, Pergamon Press, Oxford, England.

Inkret, W. C., D. Lewis, T. T. Little, G. Miller, and M. E. Schillaci, 1998a, Los Alamos National 

Laboratory Radiological Dose Assessment – Tritium Internal Dosimetry and Bioassay 

Programs, LA-UR-99-832, University of California, Los Alamos National Laboratory, Los 

Alamos, New Mexico, October 28. [SRDB Ref ID: 13485]

Inkret, W. C., D. Lewis, T. T. Little, G. Miller, and M. E. Schillaci, 1998b, Los Alamos National 

Laboratory Radiological Dose Assessment – Plutonium Internal Dosimetry and Bioassay 

Programs, LA-UR-99-834, University of California, Los Alamos National Laboratory, Los 

Alamos, New Mexico, October 28. [SRDB Ref ID: 12900]

Inkret, W. C., D. Lewis, T. T. Little, G. Miller, and M. E. Schillaci, 1998c, Los Alamos National 

Laboratory Radiological Dose Assessment – Internal Dosimetry Programs, LA-UR-99-835, 

University of California, Los Alamos National Laboratory, Los Alamos, New Mexico, October 

28. [SRDB Ref ID: 13486]

Inkret, W. C., D. Lewis, G. Miller, and M. E. Schillaci, 1998d, Los Alamos National Laboratory 

Radiological Dose Assessment – Americium Internal Dosimetry and Bioassay Program, 

LA-UR-99-836, University of California, Los Alamos National Laboratory, Los Alamos, New 

Mexico, October 30. [SRDB Ref ID: 13488]

Inkret, W. C., D. W. Efurd, G. Miller, D. J. Rokop, and T. M. Benjamin, 1999a, Los Alamos National 

Laboratory Radiological Dose Assessment – Applications of Thermal Ionization Mass 

Spectrometry to the Detection of 239Pu and 240Pu Intakes, LA-UR-98-1818, University of 

California, Los Alamos National Laboratory, Los Alamos, New Mexico, February 23. [SRDB 

Ref ID: 12903]

Inkret, W. C., D. Lewis, G. Miller, and M. E. Schillaci, 1999b, Los Alamos National Laboratory 

Radiological Dose Assessment – Tritium Internal Dosimetry, LA-UR-99-838, University of 

California, Los Alamos National Laboratory, Los Alamos, New Mexico, February 23. [SRDB 

Ref ID: 12899]

Inkret, W. C., M. E. Schillaci, Y. S. Cheng, D. W. Efurd, T. T. Little, G. Miller, J. A. Musgrave, and J. R. 

Wermer, 1999c, Internal Dosimetry for Inhalation of Hafnium Tritide Aerosols, LA-UR-98-4814, 

University of California, Los Alamos National Laboratory, Los Alamos, New Mexico, August 

10. [SRDB Ref ID: 12906]

Kinderman, E. M., H. R. Schmidt, G. J. Alkire, A. Chetham-Strode, and R. Ko, 1953, The Isotopic 

Content and Specific Activity of Pile-Produced Plutonium, HW-26584, General Electric 
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 67 of 150

Company, Hanford Atomic Products Operation, Richland, Washington, January 12. [SRDB 

Ref ID: 14495]

Kirby, H. W., 1959, The Radiochemistry of Protactinium, U.S. Energy Commission Report NAS-NS 

3016, Subcommittee on Radiochemistry, National Academy of Sciences, National Research 

Council, Washington, D.C., December. [SRDB Ref ID: 36904]

Kolodney, M., 1946, Proposed Standards Health Protection of Plutonium Workers, February 15. 

[SRDB Ref ID: 12980]

Langham, W. H., S. H. Bassett, P. S. Harris, and R. F. Carter, 1980, “Distribution and Excretion of 

Plutonium Administered Intravenously to Man, LA-1151,” Health Physics, volume 38, pp. 

1031–1060. [SRDB Ref ID: 27960]

Langham, W. H., J. N. P. Lawrence, J. McClelland, and L. H. Hempelmann, 1962, “The Los Alamos 

Scientific Laboratory's Experience with Plutonium in Man,” Health Physics, volume 8, pp. 753–

760. [SRDB Ref ID: 12912]

LANL (Los Alamos National Laboratory), 1954-1957, “Urine Bioassay Data Indicating Urine Bioassay 

Analysis (1954-1957),” University of California, Los Alamos National Laboratory, Los Alamos, 

New Mexico. [SRDB Ref ID: 27951]

LANL (Los Alamos National Laboratory), 1983a, Health Physics Quarterly Progress Report, January – 

March 1983, Los Alamos, New Mexico. [SRDB Ref ID: 13567]

LANL (Los Alamos National Laboratory), 1983b, Health Physics Quarterly Progress Report, October – 

December 1983, Los Alamos, New Mexico. [SRDB Ref ID: 13569] 

LANL (Los Alamos National Laboratory), 1983c, Health Physics Quarterly Progress Report, July – 

September, Los Alamos, New Mexico. [SRDB Ref ID: 13568]

LANL (Los Alamos National Laboratory), 1984, Respiratory Protective Equipment, Health and Safety 

Manual Technical Bulletin 1203, Los Alamos, New Mexico, August. [SRDB Ref ID: 12913]

LANL (Los Alamos National Laboratory), 1994, Explanatory Notes, Health-Safety Records, re: The 

First Official Monitoring of Any Building, 1947, Los Alamos, New Mexico, April 1. [SRDB Ref 

ID: 12968]

LANL (Los Alamos National Laboratory), 2007, Sigma Complex, Materials Science and Technology 

Division Facility Focus, LALP-07-054, Los Alamos, New Mexico, Summer. [SRDB Ref ID: 

41601]

LANL (Los Alamos National Laboratory), 2004a, LANL Air-Purifying Respirators Web-based Training 

Course, http://eshtraining.lanl.gov/esh13/webtraining/24312/htmfiles/Modules/Types/Types_[1

to 24].htm, accessed March 25, 2004. [SRDB Ref ID: 14496]

LANL (Los Alamos National Laboratory), 2004b, Health Physics Checklist, 

http://eshdb.lanl.gov/~esh12/new_eshdb/Matrices, accessed April 30, 2004. [SRDB Ref ID: 

12969]

LASL (Los Alamos Scientific Laboratory), 1944a, “Health Safety Rules (for Building H & Room V-8),”

excerpt from unknown report, p. 7. [SRDB Ref ID: 27319]
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 68 of 150

LASL (Los Alamos Scientific Laboratory), 1944b, Health - Safety Report, Chemical and Metallurgical 

Division, July, 1944, LAMS-00119, Los Alamos, New Mexico. [SRDB Ref ID: 887]

LASL (Los Alamos Scientific Laboratory), 1944c, Health and Safety Report, Chemical and 

Metallurgical Division, May, 1944, LADC-844, LAMS-99, University of California, Los Alamos, 

New Mexico [SRDB Ref ID: 889]

LASL (Los Alamos Scientific Laboratory), 1945, Chemistry and Metallurgy Health Handbook of 

Radioactive Materials, August 17, 1945, LA 391, Los Alamos, New Mexico. [SRDB Ref ID: 

17156]

LASL (Los Alamos Scientific Laboratory), 1947, H-Division Progress Report, 20 October 1947 – 20 

November 1947, LAMS-651, Los Alamos, New Mexico, December 2. [SRDB Ref ID: 7885]

LASL (Los Alamos Scientific Laboratory), 1948a, H-Division Progress Report 20 November – 20 

December 1948, LAMS-828, Los Alamos, New Mexico, pp. 183_209. [SRDB Ref ID: 28618]

LASL (Los Alamos Scientific Laboratory), 1948b, H-Division Progress Report: 20 October 1948 – 20 

November 1948, Los Alamos, New Mexico, December 2, pp. 163–182. [SRDB Ref ID: 28618]

LASL (Los Alamos Scientific Laboratory), 1949a, H-Division Progress Report: 20 April 1949 – 20 May 

1949, Los Alamos, New Mexico, pp. 91–112. [SRDB Ref ID: 28619]

LASL (Los Alamos Scientific Laboratory), 1949b, Health Division Progress Report 20 March – 20 

April, 1949, LAMS-889, Los Alamos, New Mexico, May 31, pp. 68–90. [SRDB Ref ID: 28619]

LASL (Los Alamos Scientific Laboratory), 1952a, Health –Safety Rules, January 22, 1952, Los 

Alamos, New Mexico, January 22. [SRDB Ref ID: 27981]

LASL (Los Alamos Scientific Laboratory), 1952b, Health Rules at DP East, January 9, 1952, Los 

Alamos, New Mexico, January 9. [SRDB Ref ID: 27963]

LASL (Los Alamos Scientific Laboratory), 1954, General Handbook for Radiation Monitoring, 2nd

Edition, LA-1835, Los Alamos, New Mexico, September. [SRDB Ref ID: 968]

LASL (Los Alamos Scientific Laboratory), 1955, Health Division Progress Report, H-207, December 

20, 1954 – January 20, 1955, Los Alamos, New Mexico, pp. 2–37. [SRDB Ref ID: 28628]

LASL (Los Alamos Scientific Laboratory), 1959, Compilation of Office Memorandum Concerning 

LAPRE II, Ten-Site and Other Reactors, Los Alamos, New Mexico. [SRDB Ref ID: 14513]

LASL (Los Alamos Scientific Laboratory), 1979, Evaluation of Polonium-210 Exposures for 

Termination Reports, Los Alamos, New Mexico, August 1. [SRDB Ref ID: 14519]

Lawrence, J. N. P., 1967, Standard Operating Procedure for Plutonium Urinalysis Sampling, LA-3868-

SOP, University of California, Los Alamos Scientific Laboratory, Los Alamos, New Mexico, 

October 30. [SRDB Ref ID: 12916]

Lawrence, J. N. P., 1973, Standard Operating Procedure for Plutonium Urine and Fecal Sampling, 

LA-3836-SOP (Rev), University of California, Los Alamos Scientific Laboratory, Los Alamos, 

New Mexico, July. [SRDB Ref ID: 12918]
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 69 of 150

Lawrence, J.N.P., 1976, Letter to Lawrence Stafford, “Pu Urine Samples for Certain Zia Co. 

Employees,” Los Alamos, New Mexico, March 29. [SRDB Ref ID: 64203]

Lawrence, J. N. P., 1978, A History of PUQFUA, Plutonium Body Burden (Q) from Urine Assays, 

LA-7403-H, University of California, Los Alamos Scientific Laboratory, Los Alamos, New 

Mexico, October. [SRDB Ref ID: 12920]

Lawrence, J. N. P., 1984, Uranium Internal Exposure Evaluation Based on Urine Assay Data, 

LA-10246-MS, University of California, Los Alamos National Laboratory, Los Alamos, New 

Mexico, September. [SRDB Ref ID: 7971]

Lawrence, J. N. P., 1990a, Procedures for Annual Calculation of Internal Uranium Dose Equivalents, 

January 31. [SRDB Ref ID: 12926]

Lawrence, J. N. P., 1990b, Revisions to Uranium Internal Exposure Evaluation Procedures (Jan. 

1990), January 23. [SRDB Ref ID: 12930]

Lawrence, J. N. P., 1992a, “Internal Doses from Uranium Urine Assays after 1991 Data,” 

memorandum to Internal Dosimetry, Los Alamos National Laboratory, Los Alamos, New 

Mexico, March. [SRDB Ref ID: 12933]

Lawrence, J. N. P., 1992b, “Uranium Urine Assay Procedures Changes,” memorandum to Internal 

Dosimetry, Los Alamos National Laboratory, Los Alamos, New Mexico, March. [SRDB Ref ID: 

12933]

Lawrence, J. N. P., 2004, “RE: 25,” e-mail to R. Argall, Los Alamos National Laboratory, Los Alamos, 

New Mexico, August 29. [SRDB Ref ID: 28000]

Lewis, D., 2006a, “Re: LANL ID TBD questions,” e-mail to R. Argall, Los Alamos National Laboratory, 

Los Alamos, New Mexico, July 28. [SRDB Ref ID: 27999]

Lewis, D., 2006b, “Re: LANL ID TBD questions,” e-mail to R. Argall, Los Alamos National Laboratory, 

Los Alamos, New Mexico, October 4. [SRDB Ref ID: 27999]

Lewis, D., 2006c, “Re: Draft Conference Call Minutes,” e-mail to E. Brackett, Los Alamos National 

Laboratory, Los Alamos, New Mexico, February 16. [SRDB Ref ID: 27996]

Little, T. T., G. Miller, and R. Guilmette, 2003a, “Internal Dosimetry of Uranium Isotopes Using 

Bayesian Inference Methods,” Radiation Protection Dosimetry, volume 105, number 1-4, 

pp. 413–416. [SRDB Ref ID: 12938]

Little, T. T., G. Miller, and R. Guilmette, 2003b, Identifying Uranium Intakes from Bioassay Data in the 

Presence of Environmental Background, Los Alamos National Laboratory, Los Alamos, New 

Mexico, [SRDB Ref ID: 12939]

Little, T. T., G. Miller, and R. Guilmette, 2003c, “Special Dose Assessment.” [SRDB Ref ID: 55135]

McClelland, J., and M. F. Milligan, 1954, Determination of Tritium in Urine and Water, LA01645, 

University of California, Los Alamos Scientific Laboratory, Los Alamos, New Mexico, June 30. 

[SRDB Ref ID: 12940]
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 70 of 150

McKown, D. A., 1957, “1956 Annual Report, for Sigma Building, TU Building, Tuballoy Shop, M 

Building and Building 96 for Uranium – 1/1/56 thru 12/31/56,” memorandum to D. Meyer, 

University of California, Los Alamos Scientific Laboratory, Los Alamos, New Mexico, February 

26. [SRDB Ref ID: 27983]

McKown, D. A., 1958, 1957 Annual Report, for Sigma Building, TU Building, Tuballoy Shop, M 

Building and Building 96, Los Alamos Scientific Laboratory, Los Alamos, New Mexico, 

January 27. [SRDB Ref ID: 27984]

McInroy, J. F., R. L. Kathren, G. L. Voelz, and M. J. Swint, 1991, “U.S. Transuranium Registry Report 

on the 239Pu Distribution in a Human Body,” Health Physics, volume 60, number 3, pp. 307-

333. [SRDB Ref ID: 13077]

Meyer, D. D., 1954, “Health Precautions – Maintenance Work DP East,” memorandum to J. Bolton, 

Los Alamos Scientific Laboratory, Los Alamos, New Mexico, May 5. [SRDB Ref ID: 27982]

Meyer, D. D., 1964, “Health Tests for the Week Ending April 10, 1964,” memorandum to R. Fowler, 

University of California, Los Alamos Scientific Laboratory, Los Alamos, New Mexico, April 21 

[SRDB Ref ID: 13459]

Meyer, D. D., 1970, “Report on Radioactivity Release, Oct. 7, 1970, at DP West,” memorandum to H1 Files, University of California, Los Alamos Scientific Laboratory, Los Alamos, New Mexico, 

October 16. [SRDB Ref ID: 13484]

Miller, G., W. C. Inkret, M. E. Schillaci, and D. Lewis, 1999, Los Alamos National Laboratory

Radiological Dose Assessment – Plutonium Internal Dosimetry, LA-UR-99-844, Los Alamos 

National Laboratory, Los Alamos, New Mexico, February 22. [SRDB Ref ID: 12946]

McClelland, J., 1958, Analytical Procedures of the Industrial Hygiene Group, LA-1858 (2nd Edition), 

University of California, Los Alamos Scientific Laboratory, Los Alamos, New Mexico, 

September 1. [SRDB Ref ID: 12948]

Moss, W., 1990, “Evaluation of Plutonium Bioassay Data 1944 –1977,” memorandum to P. Gautier, 

Los Alamos National Laboratory, Los Alamos, New Mexico, July. [SRDB Ref ID: 12952]

Moss, W. D., E. E. Campbell, H. F. Schulte, and G. L. Teitjen, 1969, “A Study of the Variations Found 

in Plutonium Urinary Data,” Health Physics, volume 17, pp. 571–578. [SRDB Ref ID: 14529]

Nasise, J. E., 1988, Some New Techniques in Tritium Gas Handling as Applied to Metal Hydride 

Synthesis, LA-UR-88-871, University of California, Los Alamos Scientific Laboratory, Los 

Alamos, New Mexico, September 1. [SRDB Ref ID: 70273] 

NCRP (National Council on Radiation Protection and Measurements), 1988, Exposure of the 

Population in the United States and Canada from Natural Background Radiation, Report 94, 

Bethesda, Maryland.

Nickson, J. J., 1945, Report on Conference on Plutonium – May 14th and 15th, CH-3167, Los Alamos 

Scientific Laboratory, Los Alamos, New Mexico, July 28. [SRDB Ref ID: 27320]

NIOSH (National Institute for Occupational Safety and Health), 2006, SEC Petition Evaluation Report 

Petition SEC-00061, Office of Compensation Analysis and Support, Cincinnati, Ohio, 

August 18.
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 71 of 150

NIOSH (National Institute for Occupational Safety and Health), 2007a, Radiation Exposures Covered 

for Dose Reconstructions Under Part B of the Energy Employees Occupational Illness 

Compensation Program Act, OCAS-IG-003, Rev. 0, Office of Compensation Analysis and 

Support, Cincinnati, Ohio, November.

NIOSH (National Institute for Occupational Safety and Health), 2007b, SEC Petition Evaluation Report 

Petition SEC-00051, Office of Compensation Analysis and Support, Cincinnati, Ohio, 

February 1.

Oppenheimer, J. R., 1944, “Request for gas masks,” record of telegraph transmission to Brigadier 

General L. Groves, Los Alamos, New Mexico, February 3. [SRDB Ref ID: 13462]

ORAUT (Oak Ridge Associated Universities Team), 2005a, Internal Dose Overestimates for Facilities 

with Air Sampling Programs, ORAUT-OTIB-0018, Rev. 01, Oak Ridge, Tennessee, August 9.

ORAUT (Oak Ridge Associated Universities Team), 2005b, Guidance on Wound Modeling for Internal 

Dose Reconstruction, ORAUT-OTIB-0022, Rev. 00, Oak Ridge, Tennessee, November 18

ORAUT (Oak Ridge Associated Universities Team), 2005c, Savannah River Site, ORAUT-TKBS0003, Rev. 03, Oak Ridge, Tennessee, April 5.

ORAUT (Oak Ridge Associated Universities Team), 2007a, Los Alamos National Laboratory – 

Occupational External Dosimetry, ORAUT-TKBS-0010-6, Rev. 01, Oak Ridge, Tennessee, 

May 30.

ORAUT (Oak Ridge Associated Universities Team), 2007b, Calculation of Dose from Intakes of 

Special Tritium Compounds, ORAUT-OTIB-0066, Rev. 00, Oak Ridge, Tennessee, April 26.

ORAUT (Oak Ridge Associated Universities Team), 2007c, Internal Dose Reconstruction, ORAUTOTIB-0060, Rev. 00, Oak Ridge, Tennessee, February 6.

ORAUT (Oak Ridge Associated Universities Team), 2008, Estimating Doses for Plutonium Strongly 

Retained in the Lung, ORAUT-OTIB-0049, Rev. 01, Oak Ridge, Tennessee, September 26.

ORAUT (Oak Ridge Associated Universities Team), 2009a, Los Alamos National Laboratory Bioassay 

Repository Database, ORAUT-OTIB-0063, Rev. 00, Oak Ridge, Tennessee, August 24. 

ORAUT (Oak Ridge Associated Universities Team), 2009b, Internal Dosimetry Coworker Data for Los 

Alamos National Laboratory, ORAUT-OTIB-0062, Rev. 00-A (draft), Oak Ridge, Tennessee.

Penneman, R. A. and T. K. Keenan, 1960, The Radiochemistry of Americium and Curium, NAS-NS3006, National Academy of Sciences, National Research Council, University of California, Los 

Alamos, California, January 31. [SRDB Ref ID: 41106], 

PNNL (Pacific Northwest National Laboratory), 2003, Methods and Models of the Hanford Internal 

Dosimetry Program, PNNL-MA-860, Richland, Washington. [SRDB Ref ID: 11521]

Roark, Kevin, 2003, (American Nuclear Society Criticality Safety Division Newsletter), ‘Neptunium 

Criticality Achieved,’ Winter, accessed on line: 

http://ncsd.ans.org/site/newsletters/2003/winter03.pdf.” 
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 72 of 150

Schulte, H. F., and D. D. Meyer, 1957, Control of Health Hazards in Handling Plutonium, Results of 14 

Years Experience, LADC 3318 revised, Los Alamos Scientific Laboratory, Los Alamos, New 

Mexico. [SRDB Ref ID: 13463]

Shipman, T. L., 1959, H-Division Progress Report, August 20 – September 20, 1959, University of 

California, Los Alamos Scientific Laboratory, Los Alamos, New Mexico, pp. 114–127. [SRDB 

Ref ID: 28638]

Shipman, T. L., 1964, memorandum with attachment to Distribution, University of California, Los 

Alamos Scientific Laboratory, Los Alamos, New Mexico, September. [SRDB Ref ID: 12915]

Shipman, T. L., T. N. White, R. S. Grier, R. Reider, W. H. Langston, H. F. Schulte, and D. D. Meyer, 

1951, Annual Report, 1951, Health Division, University of California, Los Alamos Scientific 

Laboratory, Los Alamos, New Mexico, May. [SRDB Ref ID: 972]

Smith, J. L., J. C. Spirlet, and W. Muller, 1979, Superconducting Properties of Protactinium, Science, 

205, 188-190, July 13. [SRDB Ref ID: 41605]

Stout, E. L., 1958, Safety Considerations for Handling Plutonium, Uranium, Thorium, the Alkali Metals, 

Zirconium, Titanium, Magnesium, and Calcium, LA-2147, University of California, Los Alamos 

Scientific Laboratory, Los Alamos, New Mexico, January 17. [SRDB Ref ID: 40075] 

Tribby, J. F., 1946a, “Health Protection of Maintenance Personnel Working in Contaminated Areas,” 

memorandum to Mr. Marsh (Zia Corporation), Los Alamos Scientific Laboratory, Los Alamos, 

New Mexico, May 14. [SRDB Ref ID: 27980]

Tribby, J. F., 1946b, “Contaminated Fires,” memorandum to H. I. Section Leaders, Los Alamos 

Scientific Laboratory, Los Alamos, New Mexico, April 24. [SRDB Ref ID: 27978]

University of California, 1977a, Quarterly Progress Report, July – September 1977, Group H-1, Health 

Physics, Los Alamos Scientific Laboratory, Los Alamos, New Mexico. [SRDB Ref ID: 27314]

University of California, 1977b, Quarterly Progress Report, January – March 1977, Group H-1, Health 

Physics, Los Alamos Scientific Laboratory, Los Alamos, New Mexico. [SRDB Ref ID: 27309]

University of California, 1977c, Quarterly Progress Report, October – December 1977, Group H-1, 

Health Physics, Los Alamos Scientific Laboratory, Los Alamos, New Mexico. [SRDB Ref ID: 

27137]

University of California, 1977d, Quarterly Progress Report, April – June 1977, Group H-1, Health 

Physics, Los Alamos Scientific Laboratory, Los Alamos, New Mexico. [SRDB Ref ID: 27311]

University of California, 1978a, Quarterly Progress Report, April – June 1978 Operational 

Environmental Health Safety Activities, Los Alamos Scientific Laboratory, Los Alamos, New 

Mexico. [SRDB Ref ID: 13472]

University of California, 1978b, Quarterly Progress Report, July – September 1978, Group H-1, Health 

Physics, Los Alamos Scientific Laboratory, Los Alamos, New Mexico. [SRDB Ref ID: 13694]

University of California, 1980a, Quarterly Progress Report, January – March 1980, Group H-1, Health 

Physics, Los Alamos Scientific Laboratory, Los Alamos, New Mexico. [SRDB Ref ID: 13480]
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 73 of 150

University of California, 1980b, Quarterly Progress Report, April – June 1980, Group H-1, Health 

Physics, Los Alamos Scientific Laboratory, Los Alamos, New Mexico. [SRDB Ref ID: 13696]

University of California, 1980c, Quarterly Progress Report, July – September 1980, Operational 

Environmental Health Safety Activities, Los Alamos Scientific Laboratory, Los Alamos, New 

Mexico. [SRDB Ref ID: 13482]

University of California, 1996, Wing 9 Hot Cells Support Work Involving Highly Radioactive Materials,” 

Internet page: http://www.lanl.gov/source/orgs/nmt/nmtdo/AQarchive/99spring/wing9.html, 

June 26. 

Van Dilla, M.A., 1959, “LAPRE II Gas Leak, Radioactive Contamination of Personnel (April 22, 1959),” 

memorandum to D. Meyer, H-1, Los Alamos Scientific Laboratory, Los Alamos, New Mexico, 

April 22. [SRDB Ref ID: 14527]

Vasilik, D. G., and I. C. Aikin, 1983, In Vivo Assessment of Whole Body Radioisotope Burdens at the 

Los Alamos National Laboratory, LA-9858-MS, University of California, Los Alamos National 

Laboratory, Los Alamos, New Mexico, August. [SRDB Ref ID: 925]

Vasilik, D. G., I. C. Aikin, K. L. Coop, and C. J. Umbarger, C. J., 1984, In Vivo Assessment of Lung 

Burdens at the Los Alamos National Laboratory, LA-9979-MS, University of California, Los 

Alamos National Laboratory, Los Alamos, New Mexico, April. [SRDB Ref ID: 12960]

Vier, D. T., 1951, “HEALTH,” memorandum, Los Alamos Scientific Laboratory, Los Alamos, New 

Mexico. [SRDB Ref ID: 40656]

Voelz, G. L., R. S. Grier, and L. H. Hempelmann, 1985, “A 37-Year Medical Follow-up of Manhattan 

Project Pu Workers,” Health Physics, volume 48, number 3, pp. 249–259. [SRDB Ref ID: 

13094]

Voelz, G. L., L. H. Hempelmann, J. N. P. Lawrence, and W. D. Moss, 1979, “A 32-Year Medical 

Follow-up of Manhattan Project Plutonium Workers,” Health Physics, volume 37, pp. 445–485. 

[SRDB Ref ID: 13093]

Voelz, G. L., and J. N. P. Lawrence, 1991, “A 42-y Medical Follow-up of Manhattan Project Plutonium 

Workers,” Health Physics, volume 61, number 2, pp. 181–190. [SRDB Ref ID: 13092]

Whicker, J. J., 2004, “Relationship of Air Sample Measurements to Internal Dose: A Review,” 

Proceedings of the HPS 2004 Midyear Meeting, Health Physics Society, McLean, Virginia. 

[SRDB Ref ID: 13096]

Whicker, J. J., J. C. Rodgers, C. I. Fairchild, R. C. Scripsick, and R. C. Lopez, 1997, “Evaluation of 

Continuous Air Monitor Placement in a Plutonium Facility,” Health Physics, volume 72, 

pp. 734–743. [SRDB Ref ID: 13095]

Wingfield, E. E., 1974, “Inventory of Quantities and Locations of Radioactivity in the Environment on 

and Near AEC Sites,” memorandum to R. Taschek (Los Alamos Scientific Laboratory), U.S. 

Atomic Energy Commission, Los Alamos, New Mexico, February 19. [SRDB Ref ID: 12978]
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 74 of 150

GLOSSARY

aging

In the context of reactor fuel and mixtures of plutonium isotopes, the time since 241Am was 

separated from the plutonium mixture.

activation

The induction of radioactivity in material by irradiation with neutrons.

activity fraction

The fraction of the total activity represented by a particular radionuclide.

BiPO4 era

The period when plutonium samples were analyzed by the BiPO4 technique; October 1949 to 

January 1957.

concentration guide

The average concentration of a radionuclide in air or water to which a worker can be 

continuously exposed without exceeding acceptable radiation dose standards.

cupferron era

The period when plutonium samples were analyzed using the cupferron analytical technique; 

March 1944 to October 1949.

DTPA

Chelating agents in the form of calcium salts (CaDTPA) or zinc salts (ZnDTPA) of 

diethylenetriaminepentaacetic acid.

dry box

The predecessor to the modern-day glovebox, made of wood with ports for attached rubber 

gloves. The atmosphere often included inert gas to further contain the dust. High-efficiency 

filters on dry box exhaust and pressure differentials were used to control the spread of 

contamination to the outside. The operator could perform tasks completely separate from the 

material without exposure to the dusts, etc.

implosion

A sudden inward compression and reduction in volume.

ionium

Thorium-230, the decay product of 234U.

kiva

A remotely controlled critical assembly building associated with the Critical Experiment Facility.

nanocurie-year (nCi-yr)

The product of the current incremental body burden times the number of years from the 

estimated date of the incremental uptake to the date of the calculation. For a deceased 

person, it is the date of death; for a living person, it is the first date of the month after the latest 

sample result. Incremental nanocurie-years can be summed to provide total nanocurie-years. 
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 75 of 150

Oralloy

Enriched uranium.

rad

Unit of absorbed dose. 

simulated

In the context of urine sampling, collection of urine from about 0.5 hour before retiring to bed, 

through the sleep period, and for about 0.5 hour after rising for two consecutive nights, or 

other similar protocol, to simulate a 24-hour sample.

tolerance values

The concentration of a radionuclide in a bioassay sample above which an unacceptable intake 

had occurred or an unacceptable body burden existed in that individual.

T3O8 

Uranium oxide based on Tuballoy (natural uranium)

Tuballoy

Natural uranium.
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 76 of 150

TABLE OF CONTENTS 

SECTION TITLE

A.1 Scope.....................................................................................................................................

PAGE

78

A.2 Codes Used in Bioassay and Internal Dose Records ............................................................. 78

A.2.1 Los Alamos Bioassay Data Repository Codes............................................................78

A.3 Locations and Typical Radionuclides ..................................................................................... 80

A.4 Computer Code for Validating Plutonium Bioassay Samples ................................................. 80

A.4.1 Validations Techniques PUQFUA2.............................................................................80

A.4.2 PUQFUA Primary Validation Technique .....................................................................92

A.5 Summary of Computer Codes................................................................................................ 93

A.6 Current and Historical In Vitro Bioassay Programs................................................................. 95

A.7 Estimation of Date of Intake ................................................................................................... 95

A.8 Routine Sampling Procedures................................................................................................ 96

A.9 Summary of Analytical Techniques for Plutonium .................................................................. 96

A.10 Example Bioassay Results..................................................................................................... 96

A.11 Routine Sampling Frequency ................................................................................................. 97

A.12 Sampling Protocol for Accidental Exposures.......................................................................... 99

A.13 Summary of In Vitro and In Vivo Sensitivities ......................................................................... 99

A.14 In Vivo Bioassay Routine Frequency.................................................................................... 103

A.15 Human Experimentation....................................................................................................... 103

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 1 of 38
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 77 of 150

LIST OF TABLES 

TABLE TITLE PAGE

A-1 Identification match............................................................................................................... 78

A-2 Assessment method codes for bioassay reports................................................................... 78

A-3 In vitro collection requirements ............................................................................................. 78

A-4 In vitro collection protocols.................................................................................................... 79

A-5 Sample units.........................................................................................................................79

A-6 In vivo detector systems ....................................................................................................... 79

A-7 In vivo collection protocols .................................................................................................... 80

A-8 Current and historical locations and default respiratory absorption type................................ 81

A-9 Zia employee access to plutonium areas .............................................................................. 94

A-10 Historical in vitro bioassay .................................................................................................... 95

A-11 Current in vitro bioassay performed, 1997 ............................................................................ 95

A-12 Routine sampling procedure ................................................................................................. 96

A-13 Plutonium urinalysis–sensitivities and analytical techniques ................................................. 97

A-14 Example urine bioassay results ............................................................................................ 98

A-15 Routine plutonium sampling frequency ............................................................................... 100

A-16 Routine sampling frequency 241Am exposures .................................................................... 102

A-17 Sampling protocol for tritium ............................................................................................... 102

A-18 Uranium routine sampling protocol ..................................................................................... 102

A-19 Sampling protocol for accidental plutonium exposures, 1967.............................................. 103

A-20 Sampling protocol for accidental plutonium exposures, 1973.............................................. 105

A-21 Sampling protocol for accidental plutonium exposures, 1998.............................................. 106

A-22 Sampling protocol for accidental 241Am exposures.............................................................. 107

A-23 Uranium nonroutine sampling protocol................................................................................ 107

A-24 In vitro bioassay (except plutonium and americium) sensitivity ........................................... 108

A-25 Whole-body counting detection levels................................................................................. 109

A-26 LC and MDA values for lung counting.................................................................................. 111

A-27 In vivo bioassay routine frequency...................................................................................... 112

A-28 Amount of plutonium administered to subjects by intravenous injection 

and the date of the administration....................................................................................... 113

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 2 of 38
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 78 of 150

A.1 SCOPE

The purpose of Attachment A is to summarize the data available in Section 5 and to provide 

supporting information to discussions in the text.

A.2 CODES USED IN BIOASSAY AND INTERNAL DOSE RECORDS

A.2.1

The information for the worker in the current Los Alamos Bioassay Data Repository is provided to the 

dose reconstructor in PDF format. Each of these PDF files contains pages that define the codes that 

were used in reporting the data. These include codes that relate to the verification of the identification 

match and codes that relate to the bioassay results. These codes, as defined at the date of this

document revision, are listed in Tables A-1 through A-7. However, the dose reconstructor should 

review the codes in the file to ensure they are interpreted in accordance with current status.

Los Alamos Bioassay Data Repository Codes

Table A-1. Identification match. 

Match 

type

Match 

level Description

s 1 Social Security Number was matched.

n1b 2 First name, and entire birth date, month, day, and year were matched.

nb 2 Name and birth date were matched.

n 3 Full name match.

n1 4 The first letter of the first name was matched; for example, J. J. Smith for John J. Smith and 

the last name was matched.

nmd 4 Last name and birth month and birthday were matched.

nmy 4 Name and birth month and birth year were matched.

none 5 No match was found in the LANL personnel tables.

Table A-2. Assessment method codes for bioassay reports. 

Code Description

POR Polonium by alpha spectroscopy (May 1, 1955, to present)

LS Liquid scintillation (1981 to present)

UNAA* Uranium by DNAA (January 1, 1978, to January 1, 1992)

GB Gross beta

T3R Tritium by radiometric (1950 to 1981)

UF* Uranium by fluorophotometric determination (1950 to 1986)

UR* Uranium by radiometric (gross alpha) (January 1, 1982, to present)

TIMS Thermal ionization mass spectroscopy (January 1, 1998, to present)

Ref. LAUR-05-1942 Analysis type not available. Refer to LA-UR-05-1942 (Clark 2005).

Table A-3. In vitro collection requirements. 

Code Description

Routine Collected on a regular schedule

Special One-time-only collection for incident or follow-up

Prompt Action Samples Collected or count performed after incident

Termination Sample Collected or count performed due to termination from program or LANL

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 3 of 38
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 79 of 150

Table A-4. In vitro collection protocols. 

Type Description

Spot sample Single void collection

Simulated 24-hr Morning and evening collection for 2-d period

True 24-hr A 24-hr urine collection

500-ml sample Two single voids collected for analysis

Home drinking water A 500-ml sample of individual’s home drinking water

Spot fecal A single fecal voiding

Autopsy tissue sample Collected at autopsy – tissue type listed as sample type

Unknown Sample collection protocol not available. Refer to LA-UR-05-1942 (Clark 2005).

Blood An intravenous sample of blood; see sample size and units for additional details

Table A-5. Sample units. 

Code Description

G grams

ML milliliters

L liters

Table A-6. In vivo detector systems. 

Code Description

GELI-WHOLE BODY Whole-body counting system used in SB-14 from (approximately) August 31, 

1978, to September 6, 1979, when Ge(Li) and HPGe were moved to SB-16 and 

new phoswich system was installed in SB-16.

HUMCO II-WHOLE 

BODY

These systems were in use before 1969 for contamination screening. There are 

no numerical results in the database for counts made with either of these systems.

HPGE-THYROID Hyperpure Ge detector used for thyroid counting

ORTEC-WHOLE BODY 

ORTEC WBC

Single ORTEC coaxial HPGe detector over abdomen used as whole-body counter.

PGT-5-DET-CHEST SML (Sum Low; 300 channels) five-detector array, used January 1, 2004 to July 22, 

2004

PGT-CHEST SMH (Sum High; 1024 channels) six-detector array, amplification set to capture 0 – 3 

MeV

PGT-SKULL SML (Sum Low; 300 channels) six-detector PGT array, >= 6/20/1996 <1/1/2004

PHOS-BACK Phoswich NaI/CsI

PHOS-CHEST Phoswich NaI/CsI

PHOS-HAND Phoswich NaI/CsI

PHOS-HAND WOUND Phoswich NaI/CsI

PHOS-LIVER Phoswich NaI/CsI

PHOS-SKULL Phoswich NaI/CsI

SB14 9X5 NAI-THYROID 9 × 5 NaI(Tl) in SB-14 in SB-14 was the whole-body counting system from about 

January 1, 1969, to (approximately) August 31, 1978. LC is estimated as 0.5 × 

MDA.

SB14 9X5 NAI-WHOLE 

BODY

The 9 × 5 NaI(Tl)) in SB-14 was the whole-body counting system from about 

January 1, 1969, to (approximately) August 31, 1978. LC is estimated as 0.5 × 

MDA.

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 4 of 38
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 80 of 150

Table A-7. In vivo collection protocols. 

Type Description

BASELINE Initial count before work

NEW HIRE Initial count

RECOUNT Prior count indicates need for recount

REHIRE Individual rehired and reenrolled in in vivo monitoring

REQUEST Special count requested

ROUTINE Count performed on a regular schedule

SPECIAL One-time-only count for incident or follow-up

STANDARD Count performed on a regular schedule

TERMINATION Collected or count performed due to termination from program or LANL

TRANSFER Final count performed due to individual transferring from radiological organization

A.3 LOCATIONS AND TYPICAL RADIONUCLIDES

The historical and current typical locations for radionuclides are listed in Table A-8 (Inkret et al. 1998c; 

ENSR 2002).

A.4 COMPUTER CODE FOR VALIDATING PLUTONIUM BIOASSAY SAMPLES

Plutonium samples can be marked as invalid in the database because the results did not fit statistical 

expectations. However, the dose reconstructor might find it useful to include these sample results. 

Review of studies listing the bioassay results of the original 26 workers with intakes of plutonium 

shows considerable variability between samples in both the long and short term (Voelz et al. 1979). 

Section A.9 contains a complete listing of bioassay results for one individual. These results display 

typical variability seen in other individuals involved in the study. If the dose reconstructor chooses to 

maintain the result as invalid, the logic below will provide an explanation of how the data were 

evaluated. The information in the following sections is from Lawrence (1978).

This section is for information only. The dose reconstructor should not attempt to apply these criteria 

to data sets.

A.4.1

The PUQFUA programs were revised through PUQFUA4. Therefore, other validation protocols might 

have been in effect after 1978.

Validations Techniques PUQFUA2

1. The purpose of the validation is to determine high results to invalidate the sample.

2. Only samples of a single era could be used to validate samples from that era (e.g., samples 

analyzed by BiPO4 or cupferron could be used to validate samples of the era).

3. Samples are considered positive if they exceed the value of LEAST. If the sample passes

Test A, do not perform Test B. The tests are applied sequentially to the largest results first, 

then the second largest, etc., until all results exceeding LEAST are examined. If results are 

below LEAST, use Test C.

ATTACHMENT 5A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 5 of

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 5 of 38
Document No. ORAUT-TKBS-0010 5 Revision No. 01 Effective Date: 10/15/2009 Page 81 of 150

Table A-8. Current and historical locations and default respiratory absorption type. 

Location Start

Demolished or 

decommissioned Nuclidea Comment

Inhalation class/type

ICRP 30 ICRP 68

TA-0 None Original town site

TA-1, Original Main TA 1945 1965 active

1975 

decommissioned

All (c) (c) 

TA-1, Building D, plutonium 

chemistry and metallurgy

1943 1954 Pu-239, Pu-238,

U-238, DU,

Am-241,

Po-210,

Ba-140,

La-140

Absorption depends on matrix or 

pure

Might indicate Sr-90

Y-W

Y 

W, Y (Pu matrix)

D-W 

D 

D-W

M-S

M 

M, S

F-M 

F 

F-M

TA-1, Building D-2, contaminated 

laundry

1953 Pu-239; Pu-240; 

Pu-238; U-236; DU;

Po-210;

Ac-227;

Ra-226

Depends on the compound W-Y 

W 

D 

Y 

M

M-S 

M 

F-M 

M-S 

M

TA-1, Building D-5, Sigma vault -

storage

1965 Pu-239,

U-238

W

W

M

W

TA-1, ML Building, medical 

laboratory

Cm

Am

Processing W

W

M

M

TA-1, Building C, uranium 

machining

1943 1964 Uranium W M

TA-1, Building G, uranium and 

graphite sigma pile

1959 Uranium

Ra-226

W 

W

M 

M

TA-1, Building H and Gamma 

Building

1945 1957, 1959 Po-210

MFPd InitiatorsCs-137 contamination incident 

occurred

D 

Various 

F-M 

Variable 

TA-1, Building HT, heat treatment 

and machining

1965 NU and EU W M

TA-1, Building HT Barrel House, 

storage

1964 Pu-239,

U-238

D-W-Y 

W

F-M-S 

M

TA-1, Building M, processing and 

recovery EU

EU Processing, metallurgy, and 

recovery

W M 

TA-1, Building M-1, machining U-238 W M

TA-1, Building O 1956 Radium

Radon

Radon cooked off sources on a 

hot plate, Ra/RaBe calibration 

sources

TA-1, Building Q 1959 Radium

Radon

A spill occurred

Ra calibration sources

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 6 of 38
Document No. ORAUT-TKBS-0010 5 Revision No. 01 Effective Date: 10/15/2009 Page 82 of 150

Location Start

Demolished or 

decommissioned Nuclidea Comment Inhalation class/type

TA-1, Sigma Building 1965 NU, EU,

thorium

Casting, machining, powder 

metallurgy

W 

W-Y

M 

M-S

TA-1, Building TU, machining 

Tuballoy

1964 NU W M

TA-1, Building TU-1, recovery of EU 1964 EU Furnace for burning rags W M

TA-1, Building V, Machine Shop 1943 1959 Uranium Unusual assignments W M

TA-1, Building W, Van de Graaff 

accelerator

Uranium

Po-210

tritium

Th-228 “Mesiothorium” Ra-228

W

D 

D 

--

M,

F M

FSR-2 

M-S

TA-1, Building X, Cyclotron Beryllium, uranium, 

lithium, tritium, 

strontium targets.

Zn-65

Targets had induced beta activity. Various Various

TA-1, Building Y, Physics 

Laboratory

Tritium,

uranium

F 

M

TA-1, Building Z, Cockcroft-Walton 

accelerator

None

TA-2, Omega West Site Housed critical experiments

TA-2, Water Boiler 1943 1974 U-235,

I-131, I-125,

Rb-88, Cs-137,

Xe-131, Ar-41

tritium

Pu-239

Enriched uranium fuel W

D 

D 

NG

M

SR-1,F

F

SR-D 

TA-2, Clementine 1946 1952 NU, plutonium,

I-131, I-125,

Rb-88, Cs-137,

Xe-131, Ar-41

Ruptured plutonium fuel rod, 

uranium reflectors

W 

D 

D 

M 

SR-1,F

F

SR-D

TA-2, OWR 1956 1995 U-235

I-131, I-125,

Rb-88, Cs-137,

Xe-131, Ar-41

Cr-51,

Na-24

Tc-99m

Enriched uranium fuel

I-125 production loop schedule – 

at times operated “around the 

clock”

M

SR-1,F

F 

SR-D 

F,M,S

F 

F,M

TA-3, South Mesa Site, technical 

facilities

1953 All Plutonium processing (c) (c) 

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 7 of 38
Document No. ORAUT-TKBS-0010 5 Revision No. 01 Effective Date: 10/15/2009 Page 83 of 150

Location Start

Demolished or 

decommissioned Nuclidea Comment Inhalation class/type

TA-3, Building 29, CMR (SM-29) 1951 Pu-238 Wing 9 handled irradiated 

uranium and plutonium In hot 

cells. Small quantities of uranium 

and plutonium, MFPs including 

iodines, Pu-238

Y S-M

TA-3, Building FE-19 Plutonium W M

TA-3, Building 34, Cryogenics Tritium 3,000 Ci HTO released in 1979 D SR-2

TA-3, Building 16, Van de Graaff 

accelerator

Tritium 800 Ci HTO released in 1977 D SR-2

TA-3, Building 35, press building NU, U-235 Normal metallic and oxides W M

TA-3, Building 39, Tech Shop DU W M

TA-3 SM-40, Physics Allc Incident contaminated large 

portion of building with Po-210 

through ventilation (late 1950s or 

early 1960s)

(c) (c)

TA-3, Building 66, Sigma Complex, 

metallurgy and fabrication

<1957 EU, DU powders, 

NU, plutonium, 

thorium

Plutonium processing

Normal metallic and oxide 

uranium (McKown 1958)

W M 

TA-3, Building 102, Tech Shops Uranium W M

Plutonium Y S

TA-3, Building 141, Rolling Mill DU W M

Plutonium Y S

TA-3, CMR Building, Wings 3,5,7 H-3 HTO, HT D SR-2, SR-D

TA-3, CMR Building, Wing 9 1961 Cs-137 Potential for low-level chronic 

intake in hot cell work

D F 

MFPd including

I-131

D SR-1

F

Pu-238, -239, -240 0.1-10μm AMAD, oxide, nitrate, 

fluoride, and metal. Oxide is 

most common.

Usually Y, can 

be W

S-M

TA-3, IBFa, SM-16 I-125 Iodide, labeled organics D SR-1, F

H-3 Labeled DNA precursors (OBT), 

water (HTO), HT

D SR-2(HTO, 

OBT), 

SR-0 (HT)

P-32 Labeled organics, phosphates W F-M

TA-3 Tritium Instrument Calibration 

Facility, SM-40

H-3 HTO, HT D SR-2, SR-0 

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 8 of 38
Document No. ORAUT-TKBS-0010 5 Revision No. 01 Effective Date: 10/15/2009 Page 84 of 150

Location Start

Demolished or 

decommissioned Nuclidea Comment Inhalation class/type

TA-3-184, Occupational Health Pu Y S-M

TA-3-216 Weapons Test Support Pu Y S-M

TA-3-700,

SM-700

Acid Neutralization and Pump 

Building

Pu Y S-M 

TA-4, Alpha Site 1956 DU Firing site until 1956, materials 

disposal site C

W M 

TA-6, Two-Mile Mesa Site 1944 1950 DU Detonator manufacturing W M

TA-8, Nondestructive testing 1984 Pu-239; Pu-238

U-235; DU

Gun firing site W-Y 

D-W-Y

M-S 

Co-60; Ir-192 W M

Cs-137 Y S

D F

TA-10, Building, CMR-10, Bayo 

Canyon Site

1944 1950 Sr-90, RaLa radiochemistry D or Y F or S

DU, NU W W

La-140 D or W F or M

Ba-140 D F

TA-11, K Site 1947 Ra-226-Be 20-MeV betatron

TA-15, Electron accelerator 1962 Present Pu-239 PHERMEX W M

DU W M

H-3 D SR-2

TA-16, S Site, WETF 1989 Present Pu-239 Explosive casting and machining W M

DU W M

H-3 D SR-2

TA-16, WETF, Bldg 205 Present H-3 Labeled DNA precursors (OBT), 

water (HTO), HT

D SR-2 (OBT, 

HTO)

SR-0 (HT)

TA-18 U-235; U-233

Pu-239; Pu-240;

MFPd

; 

I-131;

polonium

0.1-10 μm AMAD, oxide, nitrate, 

fluoride, and metal. Oxide is 

most common.

Ruptured polonium source, 1953

W

Usually Y, can 

be W

D-Y 

D 

D

M

M 

F-S 

SR-1, F

F

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 9 of 38
Document No. ORAUT-TKBS-0010 5 Revision No. 01 Effective Date: 10/15/2009 Page 85 of 150

Location Start

Demolished or 

decommissioned Nuclidea Comment Inhalation class/type

TA-18, Pajarito Laboratory, Rover 

reactor, criticality experiments

1946 1973 MFPsd Betatron used from 1951 to 1954,

EU metal sphere 1952; 

plutonium core added 1 year

later.

1954 unreflected, delta phase 

plutonium.

TA-19, East Gate Laboratory 1962 None

TA-21, DP-West, plutonium facility 11/45 WGPu Y M-S

TA-21, CMR, heat sources Pu-238 Accident with glovebox breach, 

1971

Y M-S

TA-21 DP West Pu-238, -239, -240 0.1-10 μm AMAD, oxide, nitrate, 

fluoride, and metal. Oxide is 

most common.

Usually Y, can 

be W

TA-21, Buildings 2 and 3, wet 

chemistry

1982 Pu 1958 accident, separated phases 

in plutonium process tank, 

unshielded tank

Y M-S

TA-21, Buildings 4 and 5, dry 

chemistry

1981 Pu Y M-S

TA-21, Building 12, filter building 1975 Pu Actinium-contaminated Y M-S

TA-21, Building 3, oxalate 

precipitation operations

Pu-239

Pu-238

U-235

Y 

Y 

W

M-S 

M-S 

M

TA-21, Building 4 1945 1948 EU hydride

TA-21, Building 4 1960 Pu-239 Hot cell examine irradiated 

plutonium and EU fuel elements

Y M-S

TA-21 1965 Pu-238 and

Pu-239

Y M-S

TA-21, Building 5, plutonium 

fabrication

Limited use in 1975 Pu-239, -238 Fire contaminated exhaust filter, 

1959

Y M-S

TA-21, Building 150, plutonium fuels 

development, heat sources 

development

1963 Pu-238, -239 Sealed capillary broke, 2,800 ×

MPC

10/1970

Y M-S

TA-21, Building 210, plutonium 

research

Plutonium Y M-S

TA-21, Building DP-East EU W M

TA-21, Building 155, TSTA, 

deuterium and tritium fuels.

1984 1990 HT and HTO >10 billion Ci. Equipment failure 

– H-3, 13.8 Ci released

D SR-2, SR-0

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 10 of 38
Document No. ORAUT-TKBS-0010 5 Revision No. 01 Effective Date: 10/15/2009 Page 86 of 150

Location Start

Demolished or 

decommissioned Nuclidea Comment Inhalation class/type

TA-21, Buildings 151, 152, 

experimental program

1945 Po-210 Produced initiators D, W F-M

Ac-227 Produced initiators Y S

TA-21, Building 155 1949 1984 Po-210 Produced initiators D, W F-M

Ac-227 Produced initiators Y S

TA-21, Building 153 1945 In service until 1970–

1973

Po-210 Produced initiators D, W F-M 

TA-21 DP-East, TSTA Tritium Test 

Assembly Facility, Building 155 

and the Salt Laboratory, Building 

209 

H-3 Labeled DNA precursors (OBT), 

water (HTO), HT

D SR-2, SR-0 

(HT)

TA-21, Liquid Waste Reprocessing, 

Buildings 35 and 257

Late 1940s 1986 Allc Plutonium and transuranic liquid 

wastes

(c) (c) 

TA-22, TD Site Ac-227 Produced initiators Y S

TA-23 NU Site 1945 1950 Unknown Firing site

TA-24 T Site 1944 DU Facilities transfer to TA-16 W M

TA-25, V Site 1944 1946 DU Taken over by TA-16

TA-26, D Site 1946 1948 U-235, U-238

H-3

Storage vault W

D

M

SR-2

TA-27, Gamma Site (Far Point) 1945 1947 Pu-239

DU

thorium

Plutonium gun assembly Usually Y, can 

be W

W

M-S 

M 

M-S

TA-28, Magazine A 1979 Present DU Firing site W M

TA-29, Magazine B 1957 DU Explosives storage area W M

TA-30 through TA-32 Unknown

TA-33, HP Site, High Pressure 

Tritium Laboratory, Building 86

1950s 1980s late H-3 Tritium oxide (HTO), tritium gas 

(HT)

D SR-2, SR-0 

TA-35, LAMPRE 1955 1967 MFPs

d Molten plutonium fuel 

Sr-90 F,S

Co-60 M,S

VFPse

MAPf

TA-35, LAPRE I, LAPRE II test 

reactors

1955 1960 MFPs

d Highly enriched U fuel

Sr-90 F,S

Co-60 M,S

VFPs

MAPsf

TA-35, Target Fabrication Facility 

(TFF), TSL-213

H-3 Labeled DNA precursors(OBT), 

water (HTO), HT

D SR-2, SR-0

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 11 of 38
Document No. ORAUT-TKBS-0010 5 Revision No. 01 Effective Date: 10/15/2009 Page 87 of 150

Location Start

Demolished or 

decommissioned Nuclidea Comment Inhalation class/type

TA-35, Ten Site, CMR-10 1950 1963 La-140, Sr-90 contamination suspected 

(F,S) 

D-W F-M

Ba-140 D F

TA-35 1981 U-235 General site W M

DU W M

Np-237 M

Plutonium M-S

Polonium F-M

TA-35, Laser Fusion Research 1974 Unknown

TA-36, Kappa Site 1950 Present DU W M

TA-37, Magazine Area C DU W M

TA-39, Ancho Canyon Site 1955 1960 NU, DU

thorium

Firing points W M

TA-40, Detonator Firing Site H-3 D SR-2, SR-0

TA-41, W Site, Weapons Group WX H-3

plutonium

uranium

americium

Engineering of nuclear 

components

Fabrication of test materials

D

Y-D 

W 

W-Y

SR-2, SR-0

F-S 

M 

M-S

TA-41, Ice House, Building 4 H-3 Tritium oxide (HTO), tritium gas 

(HT)

SR-2, SR-0

TA-42, Incinerator Site 1970 Allc Reduced low-level plutoniumcontaminated waste(c) (c) 

TA-43 HRLa 1953 1970 I-125 Iodide, labeled organics D SR-1, F

H-3 Labeled DNA precursors(OBT), 

water (HTO), HT

D SR-2, SR-0 

C-14 Labeled DNA precursors, D SR-2

P-32 Labeled organics, phosphates W F,M

TA-43 1953 1970 Allc (c) (c)

TA-45, Radioactive Liquid Waste 

Treatment Plant, WD Site

1951 1964 - operations 

ended

Allc Removed plutonium before 

discharging effluents 

(c) (c)

1960 1963 MFPsd

1967 – decom.

TA-46, WA Site 1950 1974 Rover batteries

TA-46, WA Site 1976 1980s U-235, -238 Uranium isotope separation W M

Thorium M-S

TA-48, Radiochemistry Site 1950s Present All Actinide chemistry and hot cell 

isotope production

(c) (c) 

MAPs

f

MFPsd

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 12 of 38
Document No. ORAUT-TKBS-0010 5 Revision No. 01 Effective Date: 10/15/2009 Page 88 of 150

Location Start

Demolished or 

decommissioned Nuclidea Comment Inhalation class/type

TA-48, Nuclear Chemistry 1950s Present Se-75 Spallation product, seen in hot 

chemistry on targets

M or F

H-3 Tritium oxide (HTO), tritium gas 

(HT)

D SR-2, SR-0

Cd-109 CL-/NO3

- mixture loaded in 

SnPO4 resin. Cd phosphate is 

most probable material of intake, 

very soluble, 1 μm AMAD

D F

I-131 Fission product chemistry F

TA-49, Frijoles Mesa Site 1960 1961 H-3 

plutonium

uranium

D 

Y-D 

W

SR-2, SR-0 

F-S 

M

TA-50, Waste Management Site 1963 Pu-238, -239, -240

Allc

0.1-10 μm AMAD, oxide, nitrate, 

fluoride, and metal. Oxide is 

most common.

Usually Y, can 

be W

S

TA-51, Environmental Research 

Facility

1962 Present Co-60

Sr-90

Animal exposure facility.

Presently environmental impact 

research.

Y, possibly W

D, Y

S, possibly M

F, S

TA-52, Reactor Development Site Late 1960s 1970 U-238,

Pu-238

H-3 

VFPse

krypton, xenon

High-temperature, gas-cooled, 

graphite reactor, UHTREX (UltraHigh Temperature Reactor 

Experiment)

W

Y 

D 

M

M-S 

SR-2, SR-0 

TA-53, Los Alamos Neutron Science 

Center, largest accelerator facility, 

Los Alamos Meson Physics 

Facility (LAMPF)

1972 C-11, N-13, O-15, 

Ar-41

I-131

Short-lived air activation

Medical isotope production

SR-2

SR-D

F

Induced activity in 

uranium targets, 

corrosion products

M, S

TA-54, Waste Storage Facility Pu-238, -239 -240

Allc

0.1-10 μm AMAD, oxide, nitrate, 

fluoride, and metal. Oxide is 

most common.

Usually Y, can 

be W

M, S

TA-55, Plutonium Facility (PF-4) 1969 H-3 Labeled DNA precursors (OBT), 

water (HTO), HT

D SR-2, F, SR-0 

HT

TA-55, Plutonium Facility (PF-4) 1969 Pu-238, -239, -240 0.1-10 μm AMAD, oxide, nitrate, 

fluoride, and metal. Oxide is 

most common.

Usually Y, can 

be W

M, S

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 13 of 38
Document No. ORAUT-TKBS-0010 5 Revision No. 01 Effective Date: 10/15/2009 Page 89 of 150

Location Start

Demolished or 

decommissioned Nuclidea Comment Inhalation class/type

LANSCE Be-7 2A metal, metalloid behavior, 

very reactive, occurs in virtually 

massless quantities, typically 

seen when target cells are 

opened for maintenance, usually 

in oxide form 

Y S

C-11 Byproduct at LANSCE, seen in 

workers during beam cycle

D SR-2, unless 

CO; SR-1

N-13 511 keV during beam cycle D F

a. Intakes of labeled compounds do not follow the default ICRP Publication 68 models (ICRP 1995).

b. See Table A-3.

c. All = Pu-239, Pu-240, Pu-238 (Type S-M); U-235, DU (Type M); H-3 (SR-2,F); Po-210 (Type F or M); Ac-227 (Type S); Ra-226 (Type S).

d. Cs-137.

e. VFPs = volatile fission products.

f. C-11, N-13, O-15, Ar-41, Be-7, Na-22, Na-24, Co-58, Co-57, Mn-54, Mn-52, V-48.

g. Delta phase plutonium = Pu-240 = 4.5%; also used sphere of 20.1% Pu-240 and 98.1% U-233.

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 14 of 38
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 90 of 150

4. LANL recognizes that unique situations occur when:

a. An individual urine result should be retained despite its being invalidated by the test 

procedures, and

b. An individual urine result should be invalidated despite its being validated by the test 

procedures.

5. Samples surviving Tests A, B, and C would be subjected to the PUQFUA1 primary validation 

technique.

Cupferron era Other era

LEAST

0.10 pCi

High

0.8 pCi

LEAST

0.075 pCi

High

0.4 pCi

Test A

1. If there are four or more positive samples (greater than LEAST) ±1 year – continue.

2. Select the four closest positive samples, ±1 year.

a. Find the average of the four samples.

b. Find the average of the standard deviation.

3. If the sample result is less than the average plus one-fourth the standard deviation average, 

retain the sample.

Test B – samples not passing the Test A criteria

1. If there are not three or more positive samples after the sample being tested, within the same 

sample era, the sample is validated by default.

2. Determine if one or more retained potential accident dates occurred between the sample being 

tested and the next earlier positive sample.

3. Perform the calculation of the expected excretion levels on the next three positive samples.

4. Perform the calculation using the retained potential accident date or 15 days before the 

sample date if no potential accident date is available.

RF = [I ÷ (E + I)]0.74 (A-1)

where:

RF = reduction factor

I = number of days between assumed exposure and sample

E = number of days between the sample being tested and the date of the later sample

CL = RF × UR (A-2)

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 15 of 38
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 91 of 150

and

PS = RF × U2S (A-3)

where: 

CL = calculated urine level

UR = result

PS = pseudo-standard deviation

U2S = standard deviation of the result

5. If the actual measured urine excretion level of two of the three later samples is greater than 

the calculated urine excretion rate minus one-third of the pseudo-standard deviation on the 

appropriate dates, retain the sample.

6. If two or more retained potential accident dates result in conflicting decisions, retain the 

sample.

Test C

Test C tests all samples less than LEAST in any era. Test C eliminates those low or negative urine 

results that, if left valid, would cause subsequent invalidation of high results that preceded them.

All low and negative samples are retained until the first sample exceeding LEAST is encountered.

Cupferron Era

1. If there is one sample >0.1 pCi in the set and the standard deviation is <0, the sample is 

rejected as too small.

2. If there are three or more samples >0.1 pCi in the set and the standard deviation is <0.1 pCi, 

the sample is rejected as too small.

BiPO4 Era

1. If there is one sample >0.075 pCi in the set and the standard deviation is <0, the sample is 

rejected as too small.

2. If there are three or more samples >0.075 pCi in the set and the standard deviation is <0.075 

pCi, the sample is rejected as too small.

NTA-ZnS or PHA Era

1. If there are at least two samples >0.075 pCi in the set and the standard deviation is ≤0, the 

sample is rejected as too small.

2. If there are at least five samples >0.075 pCi in the set and the standard deviation is ≤0.04, the 

sample is rejected as too small.

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 16 of 38
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 92 of 150

Surviving Samples

Samples surviving the above tests are examined in relation to the earliest date of a validated HIGH 

sample. The date of the earliest sample that exceeds the high criteria is used to test the surviving 

samples. All surviving samples dated before the first high samples are kept with the added notation of 

no highs.

Samples occurring after the first high sample

1. The average of the four retained samples greater than LEAST and closest in time to the 

sample being tested is calculated. There is no time limit within the analysis era on dates of 

samples greater than LEAST.

2. If the standard deviation exceeds this average, the sample being tested is kept; otherwise, it is 

rejected.

3. The sample can be rejected if the result is <0 or the standard deviation is <LEAST.

4. If there are not four retained samples greater than LEAST, the sample being tested is kept by 

default.

A.4.2

1. Starting with the latest sample and working toward earlier samples, successive pairs are 

examined. The later sample of each pair is used to test the validity of the earlier sample.

PUQFUA Primary Validation Technique

2. If the earlier sample is validated, it is used as the later of the next pair of samples to be tested.

3. If the earlier sample is invalidated (i.e., set equal to zero for the calculations), the later sample 

of that pair remains the later of the next pair to be tested. The sample next earlier than the 

invalidated sample becomes the one to be tested.

Uc = Ue[(De – E1)/(Dl

 – E1)]0.74 (A-4)

and

σc = σe[(De – E1)/(Dl – E1)]0.74 (A-5)

and

(Uc – Ul) > (1.282)Z1-α [(σc

2

/nc) + (σl

2

/nl )]1/2 (A-6)

where:

De = integer date of Ue

Dl = integer date of Ul

E1 = estimated integer date of intake

Uc = calculated urine result expected from Ue

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 17 of 38
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 93 of 150

Ue = earlier dated urine result

Ul = later dated urine result

σc = calculated standard deviation of Uc

σe = standard deviation of Ue

σl = standard deviation of Ul

(1.282)Z1-α = standard normal variable for (1 - α) one-sided confidence interval; α = 0.1

Hypothesis

1. Does the calculated urine result Uc exceed the measured result Ul? 

2. The Ul is permitted to exceed Uc because the larger value of Ul might have resulted from an 

additional intake between De and Dl

. 

The basic assumption is that, in the event of no additional intake, the urine level at any later 

time can be calculated from Langham’s urinary elimination equation, provided the date of 

intake has been established. (PUQFUA4 did not use Langham’s urinary equation, but rather 

the equations developed in the late 1980s or early 1990s, which take into account very-longterm elimination 20 or more years after uptake. Again, these are discussed fully in the 

description of PUQFUA4.)

3. Where nc = nl = 1, Uc does exceed Ul at the chosen level of significance, and sample Ue on 

date De is rejected as invalid.

A.5 SUMMARY OF COMPUTER CODES

The PUQSRT code provided a summary report from the PUQFUA database of:

1. Individuals with a potential accident recorded but no urine sample and individuals with a 

potential accident recorded after their latest urine sample

2. A list of individuals whose last urine sample indicated a body burden of >2 nCi

3. A list of individuals whose next-to-the-last urine sample indicated a body burden increase of 

>2 nCi

4. A list of individuals whose total body burden is >10 nCi

Tracking is available for 238Pu, 239Pu, and 242Pu body burdens. These reports were performed 

routinely.

The PUQFUA code provided calculation of 238Pu, 239Pu, and 242Pu body burdens and validation tests 

of the analytical results for the samples. Various versions were used over time as the code was 

upgraded to accommodate additional models and other features. The original version became 

operational in 1959. This program utilizes a set of power function elimination equations for the 

excretion of plutonium. Lawrence (1978) discusses the history and development of the program 

including its use to track accidents and potential accidents and validate urine samples. PUQFUA1 

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 18 of 38
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 94 of 150

and PUQFUA2 are revisions to the original code that address recognized deficiencies. When 

possible, results of the calculations have been compared with autopsy data. PUQFUA1 tended to 

overestimate by a factor of 2 to 8; this was corrected in PUQFUA2. The latest version was 

PUQFUA4.

Internal monitoring of individuals with a potential for exposure to plutonium was controlled using a 

computer code.

The Z1YRPUU code provided a method to mark Zia employees who were permitted access to 

plutonium areas based on the submission of their annual plutonium bioassay samples. This program 

was in use from 1976. Table A-9 lists the areas and exempt areas for access.

Table A-9. Zia employee access to plutonium areas (University of California 1978a).

Area

Urine sample 

within 425 d 

of entry Exempt

Job requiring respiratory protection X

Modifications or repairs on dry boxes or other highly contaminated 

equipment

X

Replacement of plutonium-contaminated filters at all sites X

Janitorial (long-term) work in plutonium operation areas X

Long-term operations (weeks) in areas of low levels of plutonium 

contamination (>1,000 dpm-60 cm2 and <10,000 dpm-60 cm2)

X

Decontamination of plutonium spills with >10,000 dpm-60 cm2 X

Work in burial pits at TA-54 when personnel contamination 

potential is moderate to high

X

Short-term jobs (2-3 d) when sizable quantities of plutonium 

(grams of Pu-238 or kilograms of Pu-239) are present in dry 

boxes (even when work is being done outside dry box)

X 

Supervisory personnel X 

(base urine sample on record)

Short-term jobs (2-3 d) in areas of CMR Building, Ten Site, TA-50, 

TA-55, TA-54, TA-18, TA-48, or TA-21, where there is little 

plutonium contamination (<1,000 dpm-60 cm2)

X

Jobs in other minimum exposure potential areas when respiratory 

protection is not required and possibility of plutonium 

contamination is minimal.

X

PUANUD (Plutonium Urine, Accident, Lung, and Wound Data) was a code to permit entry of urine 

assay, potential accident, lung count, and wound counting data into a single file that was used as 

input for the PUQFUA calculations. The software was used beginning in 1974 for 238Pu and 239Pu. 

Flags were available for the following categories.

1. Unspecified type of accident

2. Wound case with excision

3. Wound count equal to or greater than 0.2 nCi

4. High room air count if next year’s urine shows obvious increase

5. High nasal count if next year’s urine shows obvious increase

6. Nose count over 1,000 dpm

7. Other accident if next year’s urines show obvious increases

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 19 of 38
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 95 of 150

A.6 CURRENT AND HISTORICAL IN VITRO BIOASSAY PROGRAMS

Historically, the in vitro bioassay program included the nuclides listed in Table A-10. The current in 

vitro bioassay program includes the nuclides and techniques listed in Table A-11. The in vitro

bioassay samples have been taken at LANL throughout the history of the bioassay program. 

Sensitivities vary with technique and period.

Table A-10. Historical in vitro bioassay. 

Nuclide Nuclide

Tritium Iodine

Radium and daughters Fission products

Uranium Activation products

Plutonium Other alpha emitters

Polonium Other beta/gamma

Curium Americium

Strontium

Table A-11. Current in vitro bioassay performed, 1997 (Inkret et al. 1998a).

Material Analytical technique Number monitored

Tritium Liquid scintillation 139

Uranium Alpha spectroscopy 66

Plutonium Alpha spectroscopy 1,467

Plutonium TIMS 408

Americium Alpha spectroscopy 75

Strontium Liquid scintillation (gross beta counting–LSC sent to contract laboratory) 4

A.7 ESTIMATION OF DATE OF INTAKE

The following guidance from Lawrence (1978) might be relevant to an estimate of the date of 

plutonium intake, based on routine and special sampling protocol, when no intake date is recorded or 

to understand the LANL rationale for selection of an intake date listed with the bioassay results if no 

incident date is provided in the records. Otherwise, the dose reconstructor should use current models 

and best judgment in the determination of the dates for acute intakes.

1. Halfway between dates of consecutive pairs of samples if no potential accidents are recorded 

between samples.

2. One-half day before later sample of a pair if a potential accident is recorded on the date of the 

later sample.

3. The earliest potential accident date if several occur between the paired samples.

4. One-half day before the initial sample if the potential accident is recorded on the date of the 

initial sample.

5. The earliest potential accident date before the initial sample if any potential accidents occurred 

before the initial sample.

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 20 of 38
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 96 of 150

6. Fifteen days before the initial sample if no potential accidents occurred before the initial 

sample.

7. Estimated date of intake of the later sample if a pair is reassigned to be the estimated date of 

intake of the earlier sample, if the earlier sample is invalidated and a potential accident date 

occurred on the same date as the earlier sample.

A.8 ROUTINE SAMPLING PROCEDURES

Contamination of bioassay samples was a concern throughout the history of the program. Various 

sampling protocols were followed to minimize the potential for contamination, to detect levels above 

the tolerance level, and to attempt to obtain an equivalent 24-hour collection. Table A-12 lists the 

sampling protocols.

Table A-12. Routine sampling procedure. 

Years Nuclide Protocol

1944 Plutonium Collected on 24-hr urine sample in clean areas after decontamination. Working 

in plutonium areas operations – weekly, daily nasal swabs. Blood counts every 6 

wk (Hempelmann 1944).

1945–1952 Plutonium Collected 24-hr urine sample away from working environment to ensure 

contamination-free urine (Clark 2005). 

1953–1957 Plutonium Metal kit for collection of last voiding of day and first in morning for 2 consecutive 

days (Clark 2005).

1958–1967 Plutonium Four 1-pt bottle disposable kit. More closely approached volume of true 24-hr 

urine sample. Reduced potential contamination. 

1944 Polonium Collected on 24-hr urine sample in clean areas after decontamination. Working 

in polonium areas operations – weekly, daily nasal swabs. Blood counts every 6 

weeks (Hempelmann 1944).

1945 –1952 Polonium Collected 24-hr urine sample away from working environment to ensure 

contamination-free urine (Clark 2005).

1970–1975 Tritium >150 μCi/L – 1/d; 150-75 μCi/L – every 2 weeks; 74 -24 μCi/L – 1/wk (Healy 

1970).

1975–present Tritium Routine – every 2 weeks; or 2 hours after expected exposure then if <1 - no 

more samples, 1-10 μCi/L – 1 wk in next month, 10-100 μCi/L – weekly samples, 

>100 μCi/L –daily samples. 

1944–1967 Uranium Weekly urinalysis (Hempelmann 1944).

1967–present Uranium Annual, including drinking water sample since 1992 (Inkret et al. 1998c).

A.9 SUMMARY OF ANALYTICAL TECHNIQUES FOR PLUTONIUM

Various analysis techniques have been employed for plutonium over the history of the program. The 

techniques are summarized in Table A-13.

A.10 EXAMPLE BIOASSAY RESULTS

Table A-14 contains actual bioassay results from an individual whose 1959 autopsy results confirmed 

plutonium in the liver, lungs, bone, and lymph nodes. Plutonium-238/239 ratios indicative of early and 

late mixtures of plutonium were identified in the autopsy tissues. The late mixtures of plutonium were 

found in the lung. During his work at LANL, the individual had only 18 nasal swipes above 50 dpm. 

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 21 of 38
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 97 of 150

Table A-13. Plutonium urinalysis–sensitivities and analytical techniques (Clark 2005).

Period Urinalysis method

1944–1952 Plutonium extracted by cupferron chloroform, gas flow counter (1944), Simpson alpha 

counter efficiency = 50%, background = 1 cpm to 0.1 cpm (1945+). Could not differentiate 

polonium and plutonium. Selected based on potential exposure. 12% of blanks showed 

~1.4 cpm or 0.8 pCi. (1945) Assay results reflect no chemical blanks, recovery factor or 

counting geometry corrections.

1949–1957 Bi-phosphate-La fluoride coprecipitation, Simpson alpha counter, efficiency = 50%, 

background = 0.1 cpm 

1957–1963 Bi-phosphate precipitation – alkaline earth-phosphate precipitation, plutonium plated on 

stainless-steel disk. NTA counting method, 1,000-minute exposure to emulsion, 

background = 0.007 dpm.

1963–1965 Ion exchange/plutonium separated on an ion-exchange resin. Electroplated for NTA 

counting, 84% ±14% recovery

1966 ZnS counter

1967–1971 Either ZnS or alpha pulse height analysis (PHA) RAS permit measurement of Pu-238; 

background = 0.003 ±0.003 cpm 

1971–present All alpha PHA, computerized spectrometry 

9/18/80 Began Pu-242 analysis

1982–1986 Coprecipitation (alkaline earth PO4 or oxalate). Alpha spectrometry (60,000-s count time) 

or rapid Alpha Phosphor Scintillation. Counting (3,600 s)

1997 Alpha spectroscopy, Based on Class Y, 1 μm, Pu-239 (Inkret et al. 1999a) 

1997–present TIMS with ultra-trace chemistry and class-100 clean room and alpha spectroscopy 

methods. Based on Class Y, 1 μm, Pu-239, use of alpha spectroscopy allows direct 

measure of chemical efficiency and detection of Pu-238 (Inkret et al. 1999a).

(the level of significance). All the nasal swipes greater than 50 dpm were before 1948. Only one 

incident in 1955 showed a nasal swipe of approximately 28 dpm. Records of the work locations and 

results were carefully maintained (Foreman, Moss, and Langham 1960). No significant incidents or 

accidents were noted in the individual’s records. The probable intake scenario for this individual is

listed as long-term chronic or intermittent low-level inhalation. Review of the bioassay results should 

assist the dose reconstructor when reviewing the variability of other data sets.

A body burden of plutonium was potentially maintained by this individual since 1946. The bioassay 

results varied as listed.

Average room concentrations for each year are listed. When more than one room was involved, the 

range of average concentrations is listed. The period from 1946 to 1949 involved plutonium nitrate, 

plutonium oxalate, and plutonium fluorination. The period from 1955 to 1958 involved primary liquidliquid extraction of plutonium under greatly improved exposure conditions.

A.11 ROUTINE SAMPLING FREQUENCY

Routine sampling frequencies place upper bounds on the potential exposure for monitored individuals 

with all results less than detection levels. Table A-15 lists routine sampling frequencies for plutonium. 

Although sampling of individuals with the highest potential for intakes was performed from the 

beginning of the program in 1944, no specific information on the nonincident sampling program is 

available before Lawrence (1967). Table A-16 lists routine sampling frequencies for 241Am exposures; 

routine samples were not performed for 241Am before 1998. Routine and postexposure protocols are 

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 22 of 38
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 98 of 150

combined in Table A-17 for tritium. Table A-18 lists frequencies for uranium. Routine work or 

frequent entry in an area with beta/gamma emitting radionuclides currently requires annual wholebody count (LANL 2004a).

Table A-14. Example urine bioassay results (Foreman, Moss, and Langham 1960). 

Date (L/R nasal count)

Average room 

air concentration (dpm/m3) Activity (dpm/24 hr)

8/9/46 (189/320 dpm 7/29/46) 6–188 1.2

9/19/46 (149/19 9/5/46) 4.7

12/18/46 (57/68 12/10/46) 1.7

4/18/47 (164/106 dpm 

12/30/46); (102/61 dpm 

1/21/47); (91/135 dpm 4/1/47)

11–98 0.7

5/23/47 0.7

6/26/47 0.7

7/30/47 (144/40 dpm 7/7/47) 0.0

8/27/47 1.0

10/2/47 1.5

11/7/47 (120/78 dpm 10/3/47) 0.8

12/8/47 4.0

1/13/48 24–69 1.0

2/13/48 (0/59 dpm 2/10/48) 0.0

3/19/48 0.0

4/22/48 (86/3 dpm 4/26/48) 2

6/23/48 (244/72 dpm 6/10/48) 3.7

7/22/48 (72/1 dpm 7/2/48) 0.0

8/19/48 (65/0 dpm 8/2/48) 0.8

9/20/48 3.0

10/26/48 2.0

11/22/48 2.5

12/21/48 (50/38 dpm 12/1/48) 0.0

1/24/49 19–72 (Moved to uranium work)

1/31/49 2.0

7/14/49 1.2

2/8/50 0.8

9/1/50 0.1

2/28/51 0.8

9/4/51 0.7

5/19/52 0.3

12/14/52 0.0

9/4/53 0.0

6/4/54 Off uranium

6/18/54 0.0

6/8/55 (Moved back to plutonium)

8/1/55 3 1.2

8/12/55 (22/28 dpm 8/9/55) 0.6

8/19/55 0.0

9/30/55 0.7

11/14/55 0.7

ATTACHMENT 5A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 22 of 37

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 23 of 38
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 99 of 150

Date (L/R nasal count)

Average room 

air concentration (dpm/m3) Activity (dpm/24 hr)

12/27/55 0.6

2/9/56 3 0.0

4/5/56 0.7

4/30/56 0.0

6/8/56 0.4

7/20/56 0.6

8/23/56 0.5

9/25/56 0.0

10/24/56 0.0

11/23/56 0.0

12/17/56 0.1

1/31/57 4 0.23

2/28/57 0.68

4/12/57 0.22

5/14/57 0.12

6/14/57 0.11

7/15/57 0.03

8/19/57 0.00

9/20/57 0.20

10/22/57 0.21

10/31/57 0.39

11/14/57 0.51

1/10/58 4 0.00

2/21/58 0.65

3/25/58 0.51

5/7/58 0.025

6/19/58 0.55

7/30/58 0.49

9/15/58 0.79

11/28/58 0.47

A.12 SAMPLING PROTOCOL FOR ACCIDENTAL EXPOSURES

Awareness of the sampling protocol associated with exposures provides guidance for the association 

of results with intake regimes. Results corresponding to an incident can be expected to follow the 

appropriate protocols listed below. No information on the protocols is available before Lawrence 

(1967). It should be assumed that the protocol continued until the year of the next dated reference 

unless otherwise stated. Tables A-19 to A-23 list protocols for accidental exposures. 

A.13 SUMMARY OF IN VITRO AND IN VIVO SENSITIVITIES

Table A-24 lists sensitivity data for in vitro bioassay.

Tables A-25 and A-26 list sensitivity data for in vivo testing.

Monitoring was available for the liver and thyroid using the HPGe detector. In 1977 the MDA for the 

liver scan was reported to be less than 1 nCi for 235U and 241Am (University of California 1977b).

ATTACHMENT 5A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 27 of 37

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 24 of 38
Document No. ORAUT-TKBS-0010 5 Revision No. 01 Effective Date: 10/15/2009 Page 100 of 150

Table A-15. Routine plutonium sampling frequency.

Frequencya 1944 (Kolodney 1946) 1967 (Lawrence 1967) 1973 (Lawrence 1973)

1998 (Inkret et al. 1998b,c; 

LANL 2004a)

b,c

Monthly Great exposure

Quarterly Moderate exposure 1. Persons working with ≥10 g 

Pu-239 in chemical or 

metallurgical operations, inside 

or outside glovebox

2. Persons with 50% body burden 

(MPBB = 0.04 μCi Pu-239)

1. Persons working with ≥10g 

Pu-239 or ≥ 0.04g Pu-238 

(~0.6-0.7 Ci of either) inside 

or outside glovebox, or

2. Persons with body burden 

>20,000 pCi

Semiannual 

(biannual)

Slight exposure 1. Persons working with <10g of 

Pu-239 in chemical or 

metallurgical operations

2. Supervisors of quarterly 

sampled category

3. Persons with 25% burden 

(MPBB =0.04 μCi Pu-239)

1. Persons working with Pu but

≤10 g Pu-239 or ≤0.04 g 

Pu-238 in chemical or 

metallurgical operations,

2. Supervisors of the quarterly 

sampled category, or

3. Persons with >10,000 pCi 

body burden but <20,000 pCi

1. Working with ≥0.04 g (0.7 Ci) of 

Pu-238, analyzed by RAS

2. Performing chemical or 

metallurgical operations or 

maintenance on systems 

containing ≥10g of Pu-239 or 

Pu-240 and ≥0.04 g of Pu-238 

(0.6-0.8 Ci) analyzed once by 

TIMS and RAS and once by RAS

Annual 1. Other supervisory personnel

2. Persons working with sealed 

containers of plutonium

3. Other persons with casual 

encounters with plutonium who 

regularly work in areas where 

plutonium is handled

1. Supervisors, persons working 

with sealed containers of Pu,

2. Casual encounters with Pu, or

3. Working with prepared 

counting foils containing 

>20 mg of Pu-239 or 0.08 mg 

of Pu-238 (~1.3 mCi of either)

1. Routine work with operations of 

<10g (0.6 Ci) of Pu-239 or 

Pu-242 (0.6-0.8 Ci)

2. Performing maintenance on 

systems containing ≥10g of 

Pu-239 or Pu-240 (0.6-0.8 Ci) 

analyzed by TIMS and RAS

3. Working with operations of or 

performing maintenance on 

systems with <0.04 g of Pu-238

4. Line supervisors of personnel in 

semiannual categories

5. Transuranic (TRU) glovebox, bag 

outs, etc.

6. ESH-1 radiological control 

technicians (RCTs) who 

frequently enter work areas

7. All personnel with confirmed, 

measurable intakes of Pu-238 

and/or Pu-239

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 25 of 38
Document No. ORAUT-TKBS-0010 5 Revision No. 01 Effective Date: 10/15/2009 Page 101 of 150

Frequencya 1944 (Kolodney 1946) 1967 (Lawrence 1967) 1973 (Lawrence 1973)

1998 (Inkret et al. 1998b,c; 

LANL 2004a)

b,c

Initial 

(baseline)

All new hires or assigned to work 

with LANL where other persons 

are routinely sampled

1. If there has been risk of 

exposure to plutonium or if 

exposure history is missing or 

inconclusive

2. Entry requirement for unescorted 

access to some facilities

3. Occasional work in plutonium 

areas but do not handle 

plutonium

Termination 1. Persons terminating 

employment who have 

previously submitted urine 

samples, or

2. Have been working in major 

plutonium areas, but never 

sampled

Workers who submitted routine 

samples

UPPU Club

(see Section 

5.1)

Complete bioassay and physical 

examination at 5-year intervals. 

Bioassay continues after termination 

of employment.

Complete bioassay and physical 

examination at 5-year intervals. 

Bioassay continues after 

termination of employment.

Complete bioassay and physical 

examination at 5-year intervals. 

Bioassay continues after termination 

of employment.

a. The frequencies above do not apply to working with sealed sources in TA-15 where bioassay is as needed.

b. Samples analyzed by TIMS – thermal ionization mass spectroscopy.

c. Samples analyzed by RAS – radiometric, alpha-spectroscopy.

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 26 of 38
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 102 of 150

Table A-16. Routine sampling frequency 241Am exposures. 

Frequency 1998 (Inkret et al. 1998d; LANL 2004a)

a

Quarterly None

Semiannual 

(biannual)

1. Performing operations in gloveboxes with ≥0.2 g of Am-241 (0.6 Ci)

2. Performing maintenance on systems containing ≥0.2 g of Am-241 (0.6 Ci) 

Annual 1. Performing operations in gloveboxes with <0.2 g of Am-241 (0.6 Ci)

2. Frequent entry or performing maintenance on systems containing <0.2 g of Am-241 

(0.6 Ci)

3. Line supervisors of personnel in semiannual categories

4. ESH-1 RCTs who frequently enter work areas

5. All personnel with confirmed, measurable intakes of Am-241 

Initial (baseline) If there has been risk of exposure to Am or if exposure history is missing or inconclusive.

Termination Workers who submitted routine samples

a. Samples analyzed by RAS – alpha-spectroscopy.

Table A-17. Sampling protocol for tritium (Clark 2005).

Period Program Protocol

1950–1970 Biweekly for significant quantities or more often if exposure 

confirmed, removed from tritium work at 124 μCi/L.

1971–1975 >150 μCi/L – 1/d; 150-75 μCi/L – every 2 weeks;

74 -24 μCi/L – 1/wk (Healy 1970)

1975–1998 Routine – every 2 weeks; or 2 hours after expected 

exposure then if <1 μCi/L - no more samples, 1-10 μCi/L – 

1 week in next month, 10.1-100 μCi/L – weekly samples, 

>100 μCi/L –daily samples (Inkret et al. 1998a)

1998–present Work on regular or intermittent 

basis with or on systems that 

have contained 1 Ci in HT and 

any other form, or 0.1 Ci HTO, 

or MT, or 0.1 Ci of organic 

tritium

Every 2 wk 

Table A-18. Uranium routine sampling protocol (Clark 2005).

Period Protocol

Pre-1983 Biweekly samples collected.

Collected on Fridays, minimum annual

8/1983–6/1993 Spot sample collected on Monday mornings before entering work area

7/1993 Last voiding on Sunday night and second on Monday morning, first voiding

1998 Spot samples every 2 weeks/persons performing hands-on work/potential for 100 mrem; 

machining operations, polishing operations, foundry work, chemistry operations in which 

>10 g of U, work with oxidized metal, >100 g of bulk powder.a

a. Source: Inkret et al. (1998c).

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 27 of 38
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 103 of 150

Table A-19. Sampling protocol for accidental plutonium exposures, 1967 (Lawrence 1967).

Severity class Description Sampling protocol

PA (Prompt Action) Most serious accidents

1. Injection detectable by plutonium wound 

monitor (~0.01 μg)

2. Chemical burns from plutonium solutions

3. Facial contamination >20,000 cpm

4. Nose swipes >500 dpm

1. Evening after accident

2. 10 day later

3. 1 month later

4. 14-monthly thereafter.

DA (Delayed Action) 1. No known equipment failure, 10 × MPC 

for 1 week or 50 × MPC for 1 day

2. Skin contamination in excess of limits

3. Superficially contaminated cuts that are 

positive on surface monitoring only.

4. Equipment failure causes exposure for 

indeterminate period.

1. 1 month after accident

11 more samples at monthly 

intervals. No entry is made in “No. 

of days accident prior to sample” 

column of PUQFUA.

NRS (Nonroutinely 

Sampled)

1. Small wounds occurring in major 

plutonium areas (DP-West, Ten Site, 

CMR Building)

2. Exposures without respirators at 10 × 

MPC for period less than 1 week

3. Nose counts >50 dpm

4. Skin contamination >500 cpm/60 cm2 

1. 1 urine sample collected 1 month 

after intake. 

Automatic 

Rescheduling

1. Detect unexpected high exposure and 

verify its existence

2. Detect contaminated urine sample

3. Previous sample is 1 dpm/sample

1. Request another sample

2. Persons routinely excreting 

plutonium are exempt from special 

sampling but are maintained on 

routine sampling

A.14 IN VIVO BIOASSAY ROUTINE FREQUENCY

Table A-27 provides the standard protocol for scheduling in vivo counts. This is the current protocol. 

No information is available on when the protocol was first established; however, it has been in effect 

as long as anyone currently involved in the program can remember.

A.15 HUMAN EXPERIMENTATION 

The first research into the behavior and deposition of plutonium in the human body began by accident 

in August 1944 when a vial containing 10 mg of plutonium exploded in the face of a chemist, who 

swallowed a portion of the material. The chemist then used his own urine to develop a bioassay 

analysis to detect the early analysis technique for plutonium. However, this information did not 

provide excretion rates and deposition patterns necessary to assess the health risks for other workers 

appropriately.

The Plutonium Experiment involving plutonium tracers was undertaken between April 1945 and July 

1947. This was a joint project between LANL and the Atomic Energy Project of the University of 

Rochester School of Medicine and Dentistry. None of the subjects was administered the injections of 

plutonium at Los Alamos. However, the data were collected and analyzed by Wright Langham and 

others at Los Alamos. The subjects of the experiments were males over 45 years of age, an 18-yearold female, an 18-year-old male, and a 5-year-old Australian male child. The child was also injected 

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 28 of 38
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 104 of 150

with radiostrontium and radiocerium. These individuals were selected because they suffered from 

terminal illness and were not likely to live beyond 10 years. Six of the individuals outlived the 

predictions including one person who died 44 years later at the age of 80 (Hughes 2000). Urine and 

fecal samples were collected from the living subjects and autopsy samples were collected from three 

patients who died within 30 days (Clark 2005). Intake results are available for 12 subjects as reported 

in Langham et al. (1980). [The original paper was published in 1950 but not declassified until 1971.] 

These intakes are listed in Table A-28 of Attachment A. However, it appears that an additional six 

subjects were also involved. These are designated as Chi-X (University of Chicago) or Cal-X 

(University of California, San Francisco, where “X” is the subject number for that location (Hughes 

2000).

It was determined that deposition was primarily in the trabecular bone, bone marrow, and liver (Clark 

2005).

Other human experimentation was conducted at Los Alamos in later years. Tritium experiments were 

conducted in the 1940s and 1950s. These experiments involved exposures of humans to HT, HTO, 

and HTO inhalation, ingestion, and submersion in vapor. Researchers used themselves and probably 

family members as subjects (Hughes 2000).

In the late 1950s, Los Alamos biomedical researchers orally administered 125I and 131I to a group of 19 

subjects. The cohort included eight children between 4 and 10 years old, as well as three 

adolescents, aged 12, 13, and 14. The remainder of the cohort consisted of eight adults whose ages 

ranged from 26 to 46. All of the subjects lived in Los Alamos and were either employed by the 

Laboratory or relatives of employees (Hughes 2000). Each subject was administered 0.01 μCi of 

iodine orally.

In 1961, research on 137Cs was conducted with a cohort of four men, aged 27, 33, 35, and 37. In the 

1960s, four human subjects participated in a study involving 65Zn with a single oral administration. 

The subjects were the researchers, a coworker, and a family member of one of the scientists. There 

were 11 types of experiments involving approximately 130 subjects (Hughes 2000).

Table A-28 lists the amounts of spectrographically pure plutonium metal injected in the 12 subjects. 

The subject designation might be identified in the telephone interview.

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 29 of 38
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 105 of 150

Table A-20. Sampling protocol for accidental plutonium exposures, 1973 (Lawrence 1973).

Severity class Description Sampling protocol

PA (Prompt Action) Most serious accidents

1. Injection detectable by plutonium wound 

monitor >2000 pCi (DL 600 pCi Pu-239 

and 200 pCi Pu-238)

2. Chemical burns from plutonium solutions

3. Facial contamination >40,000 dpm

4. Nose swipes >500 dpm

1. Evening after accident

2. 10 days later

3. 1 month later

4. 12 monthly thereafter.

Fecal samples might be collected 

DA (Delayed Action) 1. No respiratory protection, 10 × MPC for 1 

week or 50 × MPC for 1 day

2. Skin contamination after 

decontamination in excess of limits

3. Superficially contaminated cuts that are 

positive (500 dpm) on surface monitoring 

only. 

1. Fecal collected on day 2 for 

inhalation

2. 1 month after accident fecal and 

urine

3. 11 more urine samples at monthly 

intervals. Possible fecal

Potential accident entry is made in 

PUQFUA

NRS (Nonroutinely 

Sampled)

1. Small wounds occurring in major 

plutonium areas (DP-West, Ten Site, 

CMR Building) no activity detected. <500 

dpm

2. Exposures without respirators at 10 × 

MPC for a period less than 1 week

3. Nose counts >50 dpm

4. Skin contamination >1,000 dpm/60 cm2 

5. Potential accident date is noted in 

PUQFUA and no sample has ever been 

submitted or routine is not scheduled for 

at least 3 months.

1. 1 urine sample collected 1 month 

after intake.

2. Fecal samples for types 2 and 3.

3. Memorandum sent to H-1; 

potential accident entry in 

PUQFUA might be required

Automatic 

Rescheduling

1. Detect unexpected high exposure and 

verify its existence

2. Detect contaminated urine sample

3. Previous sample is 1 dpm/sample

1. Request another sample

2. Persons routinely excreting 

plutonium are exempt from special 

sampling but are maintained on 

routine sampling

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 30 of 38
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 106 of 150

Table A-21. Sampling protocol for accidental plutonium exposures, 1998 (Inkret et al. 1998c).

Severity class Description Sampling protocol

PA (Prompt Action) Most serious accidents

1. Injection detectable by plutonium wound 

monitor >0.2 nCi

2. Chemical burns from plutonium solutions, 

skin contamination >500 dpm alpha by 

60-cm2 probe or >0.2 nCi in wound area

3. Facial contamination >10,000 dpm alpha

4. Nasal >100 dpm

5. Airborne >200 DAC-hr without respiratory 

protection.

6. Skin after decontamination is >1,000 dpm 

alpha by 60-cm2 probe

7. As requested 

Pu-239, Pu-240

1. Urine samples days 1, 3, 5 after 

intake

2. Fecal samples days 1, 3, 5 –

optional

3. In vivo chest count days 3, 5 –

optional

Pu-238

4. Urine samples on days, 1, 3, 5, 8, 

16, 30, 60, 120, 240 after intake

5. Fecal samples on days 1, 3, 5 –

optional

6. In vivo chest count on days 3, 5 –

optional

Urine analyzed by RAS and TIMS

DA (Delayed 

Action)

Effective dose equivalent >100 mrem; 

committed dose equivalent >1 rem any tissue 

or organ

1. Present in room when CAM alarms

2. Other individuals on PA

3. Positive wound count

4. Chemical burns; after decontamination 

skin >100 dpm

5. Facial contamination, before 

decontamination, >1,000 dpm

6. Nasal swipes; 1 nostril >15 dpm, both 

sum = 35 dpm alpha

7. Average ≥ 40 DAC-hr, without respiratory

8. Skin contaminate >100 dpm after 

decontamination.

9. Wounds contaminated with >250 dpm (by 

alpha probe)

10. Request by manager, group leader, or 

medical

Typically 1 urine, by RAS and TIMS, if 

Pu-239, or

At direction of ESH-1 personnel.

Chelation therapy Accident scenario exceeds PA by 20-40 

times.

Medical procedure. Interpretation of early 

radiological results under advice of ESH-12 

dose assessment.

Collection for duration of chelation 

therapy, up to 60-100 days after 

therapy.

PA schedule followed by monthly 

urine samplesa

. 

Might include blood and fecal.

Or as directed.

Follow-up Any positive routine urine As directed.

a. Samples taken for up to 100 days after termination of the therapy should not be used in the calculation of intake.

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 31 of 38
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 107 of 150

Table A-22. Sampling protocol for accidental 241Am exposures (Inkret et al. 1998d).

Severity class Description Sampling protocol

Pa (Prompt Action) Most serious accidents

1. Injection detectable by americium wound 

monitor >0.2 nCi

2. Chemical burns from americium-bearing 

solutions, skin contamination >500 dpm 

alpha by 60 cm2 probe or >0.2 nCi in 

wound area

3. Facial contamination >10,000 dpm alpha

4. Nasal >100 dpm

5. Airborne >200 DAC-hr without respiratory 

protection.

6. Skin after decontamination is >1,000 dpm 

alpha by 60-cm2 probe

7. As requested 

1. Urine samples days 1,3,5, 

8,16,30,60,120,240 after 

intake

2. Fecal samples days 1,3,5 – 

optional

3. In vivo chest count days 3,5 – 

optional

DA (Delayed Action) Effective dose equivalent >100 mrem; 

committed dose equivalent >1 rem any tissue 

or organ

1. Present in room when CAM alarms

2. Other individuals on PA

3. Positive wound count

4. Chemical burns; after decontamination skin 

>100 dpm

5. Facial contamination, before 

decontamination, >1,000 dpm

Nasal swipes; 1 nostril >15 dpm, both sum 

= 35 dpm alpha

6. Average ≥40 DAC-hr without respiratory

7. Skin contaminated >100 dpm after 

decontamination

8. Wounds contaminated with >250 dpm (by 

alpha probe)

9. Request by manager, group leader, or 

medical

Typically 1 urine, by RAS

At direction of ESH-1 personnel.

Chelation therapy Accident scenario exceeds PA by 20-40 times.

Medical procedure. Interpretation of early 

radiological results under advice of ESH-12 

dose assessment.

Collection for duration of chelation 

therapy, up to 60-100 days after 

therapy.

PA schedule followed by monthly 

urine samples.

Might include blood and fecal

or as directed.

Table A-23. Uranium nonroutine sampling protocol. 

Years Nuclide Protocol

Pending input from LANL

1998 Uranium Days 1, 4, 8 after possible incidents [high airborne, high alpha skin contamination with 

>10,000 dpm or nasal swipes over 100 dpm (summed)].a

a. Source: Inkret et al. (1998b).

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 32 of 38
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 108 of 150

Table A-24. In vitro bioassay (except plutonium and americium) sensitivity. See 

corresponding sections in the document for references. 

Radionuclide Period Typea MDA Reporting level

Plutonium U/F See Table 5-9

Americium U/F See Table 5-10

H-3 1950–1968 U (1 μCi/L)b

1969c–1987 U (0.02 μCi/L)b (1 μCi/L)b,d

1988–1998 U (0.01 μCi/L) (0.1 μCi/L)d

1999–present U (0.005 μCi/L) (0.1 μCi/L)d

Uranium See Table 5-14

Isotopic uranium See Table 5-14

Fission product 1950–1970 U 50 to 100 dpm/L

Sr-90 No information availablec

Cs-137 1965–present U 100 pCi/L

Po-210 1954 U 10 dpm/L

1955–1960 U 0.1 pCi/L

P-32 1975 U 40 pCi/L.

Th-230 as thorium 1958–1963 U 0.01 μg/

1963-?–present U 20 μg/L

Pa-231 1985 U 0.88 dpm/24-hr sample

a. U = urine; F = fecal.

b. Source: Gautier (1983).

c. Currently sent to an offsite laboratory.

d. Source: Clark (2005); use the reporting level because results were not reported below this level.

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 33 of 38
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 109 of 150

Table A-25. Whole-body counting detection levels (nCi)a 

(Vasilik and Aiken 1983; Ennis 2003).

Period Nuclide LC MDAb,c 

1955-1958d Cs-137 8

Sr-90e 30

1958-1970 d Cs-137 4

Sr-90e 30

1970-1984 Be-7 0.9 1.8

Cs-134 0.9 1.8

Cs-137 0.9 2.1

Co-57 2.1 4.8

Co-60 0.78 1.8

1984 to 1998 Tl-202 0.5 0.9

C-11 (based on 511 keV)f 0.3 0.5

Eu-152 2.2 3.3

Co-58 0.5 0.9

Co-56 0.5 0.9

Hg-197 3.1 4.6

Hg-195 2.5 3.7

Hg-195m 1.8 3.2

Hg-197m 3.8 6.0

Hg-203 0.8 1.2

Hg-193m 0.7 1.5

Cs-134 0.5 1.1

Os-185 0.6 1.1

V-45 0.5 0.8

Be-7 3.4 8.7

Sc-46 0.5 0.9

Mn-54 0.5 0.9

Cs-137 0.6 1.1

Co-60 0.5 0.8

Br-77 1.7 3.4

Sb-124 0.4 0.8

Ce-141 2.2 4.4

Ce-144 12.1 24.2

Cr-51 6.4 12.8

Co-57 1.4 2.8

Cu-67 1.5 3.0

Fe-59 1.2 2.4

Se-75 1.1 2.2

Se-73 0.4 0.8

Na-22 0.4 0.8

Zn-65 0.8 1.6

1998 to presentg 511 keV 1.15 2.3

Be-7 4.2 8.4

Ce-141 1.35 2.7

Ce-144 6.5 13.0

Co-56 0.55 1.1

Co-58 0.5 1.0

Co-60 0.45 0.9

Cr-51 4.45 8.9

Cs-134 0.5 1.0

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 34 of 38
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 110 of 150

Period Nuclide LC MDAb,c 

Cs-137 0.6 1.2

Cu-67 1.5 3.0

Eu-152 1.3 2.6

Hg-203 0.6 1.2

Mn-54 0.4 0.8

Na-22 0.45 0.9

Os-185 0.45 0.9

Ra-226 20.5 41.0

Sb-124 0.45 0.9

Sc-46 0.7 1.4

Se-75 0.85 1.7

Tl-202 0.5 1.0

U-235 1.35 2.7

V-48 0.4 0.8

Zn-65 1.1 2.2

Zr-95 0.8 1.6

a. Listing of an MDA for a given radionuclide does not necessarily mean 

that the radionuclide was frequently encountered. If smaller MDAs 

are listed in the results for a given count, use those.

b. Based on 95% confidence of detection.

c. MDA = LC × 2, unless otherwise specified.

d. The HUMCO I and II systems were designed for screening subjects. 

Subjects found to have contamination levels above background were 

referred to the 8- × 4-in. NaI detector, which had the same 

sensitivities with an extended count time.

e. By bremsstrahlung.

f. C-11 is a positron emitter with no photons. However, the 511- keV 

peak should always be present due to positron annihilation. The 

511-keV peak can have interference contributions from other 

sources, including pair production interactions from nuclides with 

photon energies greater than 1022 keV.

g. Lower sensitivities might be available using the lung counter for 

certain nuclides if lung counting is appropriate to the dose 

reconstruction.

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 35 of 38
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 111 of 150

Table A-26. LC and MDA values for lung counting. 

Period Radionuclide LC MDAa

1977 – 60-min count special UPPU group

(University of California 1977b)

Am-241 0.3

Pu-238 10

Pu-239 21

1980 

(Ennis 2003)b

Am-241 0.155 0.31

Pu-238 11 22

Pu-239 24 48

1984 

(Vasilik et al. 1984)c

Am-241 0.16 0.32

Pu-238 14 28

Pu-239 30 60

1998d to present

(Ennis 2003)

Am-241 0.1 0.2

Am-243 0.1 0.2

Pu-238 10 20.0

Pu-239 31 62.0

Th-234 0.85 1.7

U-235 0.1 0.2

Np-237 0.2 0.4

Np-239 0.1 0.2

1998 to present

Fission/activation products (Ennis 2003)

511 keV 0.1 0.2

Be-7 0.35 0.7

Ce-141 0.1 0.2

Ce-144 0.25 0.5

Co-56 0.1 0.2

Co-58 0.05 0.1

Co-60 0.1 0.2

Cr-51 0.35 0.7

Cs-134 0.05 0.1

Cs-137 0.1 0.2

Cu-67 0.1 0.2

Eu-152 0.1 0.2

Hg-203 0.05 0.1

Mn-54 0.1 0.2

Na-22 0.1 0.2

Nd-147 0.1 0.2

Os-185 0.05 0.1

Ra-226 0.9 1.8

Sb-124 0.05 0.1

Sc-46 0.1 0.2

Se-75 0.05 0.1

Tl-202 0.05 0.1

a. Assume chest wall thickness of 2.3 cm; the MDA should be increased by at least 50% for very 

large individuals (University of California 1977b).

b. Assume chest wall thickness of 2.5 cm.

c. MDA = LC × 2 

d. Lung counter has 10-300 keV and 80-3,000 keV ranges, therefore, a lower sensitivity for 

certain fission and activation products can be obtained.

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 36 of 38
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 112 of 150

Table A-27. In vivo bioassay routine frequency (Inkret et al. 1998c).

Description Frequency Comments

Baseline As requested Unescorted access into radiological controlled areas, radiological buffer areas, 

or radiological areas. Personnel who occasionally work or visit areas where 

plutonium is handled but do not handle plutonium.

Pu2 Annually Routine work or system maintenance with ≥0.2 g pure Am-241.

Routine work or system maintenance with ≤0.2 g pure americium or frequent 

entry where work is performed with any pure Am-241.

Routine work or frequent entry in any area with beta/gamma-emitting 

radionuclides.

Pu1 Biennially Works with plutonium

Routine work or frequent entry in any area with Pu-239 or -240, including 

material processing, system maintenance, supervision, or other support.

Routine work or frequent entry in any area with Pu-238.

Personnel who routinely work with plutonium and perform chemical or 

metallurgical operations with <10 g of Pu-239 or -242 or 0.04 g of Pu-238.

TRU glovebox/fume hood operations

TRU hot jobs

TRU bag outs

Glovebox operations

Fume hood operations

Hot jobs

Bag outs

Maintenance operations on process systems, and

RCT duties.

Pu Semiannually Special request

Uranium Annual Weekly access to areas where DU is machined or polished, casting and 

cleaning crucibles outside dry boxes, chemical operations, including 

purification and recovery. Handling ≥100 g metal or bulk powder outside 

gloveboxes, or operations with uranium-hexafluoride in uncontained systems.

MFPs Semiannually Special request

MAPs/MFPs Annual Works in high contamination areas and/or in airborne radioactivity area.

Th Annual Works with thorium

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 37 of 38
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 113 of 150

Table A-28. Amount of plutonium administered to 

subjects by intravenous injectiona and the date of 

the administration (Langham et al. 1980).

Designation 

of the subject

Date of 

administration

μg of Pu 

injectedb

HP-1 October 16, 1945 4.6

HP-2 October 23, 1945 5.1

HP-3 November 27, 1945 4.9

HP-4 November 27, 1945 4.9

HP-5 November 30, 1945 5.1

HP-6 February 1, 1946 5.3

HP-7 February 8, 1946 6.3

HP-8 March 9, 1946 6.5

HP-9 April 3, 1946 6.3

HP-10 July 16, 1946 6.1

HP-11 February 20, 1946 6.5

HP-12 April 10, 1945 4.7

a. Plutonium was administered as Pu+4-citrate in 0.5 ml of 

0.41% sodium citrate 2H2O.

b. Average standard deviation of determination of dose 

was 3%.

ATTACHMENT 5A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 37 of 37

ATTACHMENT A

OCCUPATIONAL INTERNAL DOSE FOR MONITORED WORKERS

Page 38 of 38
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 114 of 150

Chronic intakes, or frequent intermittent intakes, can be modeled as chronic. Very low-level, frequent, 

intermittent intakes might have occurred for the highest-risk workers, but the intakes were below 

detectability at the time. For radionuclides with long residence times in the body, chronic intakes lead 

to a slow buildup of activity and a concomitant increase in urinary excretion. Studies have shown that 

room air concentration is not necessarily a reliable quantitative predictor of intake because of the 

variations in respiratory conditions and particle dispersion in an area. Even breathing zone samples 

are not always an accurate predictor of the amount of intake (Whicker 2004). However, maximum 

and average airborne concentrations can be helpful in the establishment of boundary conditions for 

intakes. Maximum and average airborne contamination levels in buildings with high exposure 

potential are listed in Table B-1. Airborne concentrations were not available for all years of operation, 

especially for fission products. Average concentrations are calculated from general air samples as 

simple averages or averages obtained directly from LANL reports. Blanks in the table indicate that no 

information is available for that period.

Table B-1. Airborne concentrations from selected buildings.

Building Nuclidea,b Year

Maximum

airborne concentrationc

dpm/m3

Average

airborne concentrationc 

dpm/m3

D Alpha 1945 674.0

D Alpha U 1945 2,564.400

H Alpha Po 1945 48.00

DP East Alpha (EU) 1945 2,458.0 68.2

DP West Alpha 1945 254.0

D Alpha (Pu) 1945 480.0 151.0

D Alpha (Pu) 1946 2,590.0 46.86

DP East Alpha (EU) 1946 147,400.0 (Po) 271.27

TU Building Alpha 1946 1,958.40 43.40

DP West Alpha 1946 2,400.0

TU Building Alpha 1947 1,357.20 185.40

H Alpha Po 1947 8.80

Sigma Total alpha U 1947 187.20

D U 1947 55.00

D Pu 1947 134.00

D Cold lab alpha 1947 15.0

Sigma U 1948 1,393.80 124.60

U Pu 1948 379.40 66.67

H Po 1948 52.80 4.50

D U 1948 2,564.40 21.23

D Pu 1948 860.80 11.53

D Cold lab alpha 1948 882.4 (normally 35.8) 5.675

Am Laboratory Am alpha 1948 6.2 1.6

Sigma Total alpha U 1948 1,393.80 124.60

DP West Alpha 1949 1,694.0

Am Laboratory Am alpha 1949 18.0

Press Room Alpha 1949 198.0

TU Building Furnace Alpha 1949 4,095.0

Waste Treatment Lab Alpha 1949 13.8

M Alpha 1949 580.8 84.0

ATTACHMENT B

AIR CONCENTRATIONS IN SELECTED BUILDINGS

Page 1 of 5 
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 115 of 150

Building Nuclidea,b Year

Maximum

airborne concentrationc

dpm/m3

Average

airborne concentrationc 

dpm/m3

DP East Po 1949 2,344.0

U Alpha 1949 37.2

D Pu 1952 2.4

CMR Pu 1952 1.9

M U 1952 4.6

D Pu 1953 600.4 3.2

CMR Pu 1953 388.7 1.0

U Pu 1953 4.5 1.4

M U 1953 210.6 2.4

Dd U 1953 1,969.3 4.3

CMR U 1953 21.1 1.2

CMR Pu 1954 2,851.3 1.3

CMR Pu 1955 527.2 0.7

CMR Pu 1956 162.7 0.7

CMR Pu 1957 351.0 0.8

CMR Pu 1958 1,370.2 1.0

CMR Pu 1959 6,712.9 1.0

CMR Pu 1960 518.1 0.9

CMR Pu 1961 426.0 1.0

CMR Pu 1962 4,680.0 1.0

CMR Pu 1963 166.0 1.0

CMR Pu 1963 4.0 0.0

CMR Pu 1967 285.0 1.0

CMR Pu 1968 5,370.0 1.0

CMR Pu 1968 11.0 0.0

CMR Pu 1969 685.0 1.0

CMR Pu 1969 11.0 0.3

CMR U 1954 77.4 0.8

CMR U 1955 112.1 0.7

CMR U 1956 67.8 1.0

CMR U 1957 2,231.1 1.4

CMR U 1958 51.3 1.3

CMR U 1959 24.0 1.2

CMR U 1960 94.1 1.2

CMR U 1961 35.0 1.0

CMR U 1962 43.0 1.0

CMR U 1963 65.0 1.0

CMR U 1963 11.0 0.0

CMR U 1967 80.0 1.0

CMR U 1968 53.0 1.0

CMR U 1968 14.0 0.6

CMR U 1969 65.0 1.0

CMR U 1969 258.0 2.4

Laundry Total alpha 1948 844 450.533

ATTACHMENT B

AIR CONCENTRATIONS IN SELECTED BUILDINGS

Page 2 of 5 
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 116 of 150

Building Nuclidea,b Year

Maximum

airborne concentrationc

dpm/m3

Average

airborne concentrationc 

dpm/m3

Laundry Total alpha 1949 2,268 583.067

Laundry Total alpha 1950 78 42.2857

TA-21e I-131 1961 1.4E4 NA

DP West I-131 1964 302 5.5

Sigma Beta -gammaf 1947 1.82E+06

H Beta -gammaf 1947 0.0

U Beta -gammaf 1947 5.95E+05

Sigma Beta -gammaf 1948 1.86E+09 8.21E+07

H Beta -gammaf 1948 3.33E+08 7.39E+07

U Beta -gammaf 1948 1.47E+09 1.95E+08

CMR MFPsf 1961 78.0 12.8

CMR MFPsf 1962 11,627.0g 19.7g

CMR MFPsf 1963 518.0 15.0

CMR MFPsf 1964 19,256.0h 39.0h

CMR MFPsf 1965 13,404.0i 22.0i

CMR MFPsf 1966 93,887.0j 366.7j

CMR MFPsf 1967 19,256.0 39.0

CMR MFPsf 1968 13,404.0 22.0

CMR MFPsf 1969 93,887.0 366.7

CMR MFPsf 1970 14,163

CMR MFPsf 1971 18,104

a. Guidance on adjusting for different enrichments of uranium or mixtures of plutonium is in Section 5.2.

b. Information on specific mixtures of fission products is not readily available from LANL Guidance.

c. Results before January 1, 1953, were reported as cpm/L. Listed results have been converted to dpm/m3 assuming 

a nominal counting efficiency of 50%.

d. D Building was demolished in 1954.

e. Dummer (1961a,b). 

f. MFPs; assumes gross beta/gamma tolerance is 1 × 10-7 μCi/cm3 (2.22 × 105 dpm/m3 ) through 1948; the MAC is 

6.7 × 103 dpm/m3 from 1961 to 1969.

g. Highest concentration of 11,627 dpm/m3 is in "uranium cell corridor - no personnel exposure.” Next highest 

concentration is 207 dpm/m3

. Average concentration would be 11.8 dpm/m3. 

h. "No personnel exposure" to highest concentration of 19,256 dpm/m3

. Next highest concentration is 2,551 dpm/m3. 

i. "No personnel exposure" to highest concentration of 13,404 dpm/m3

. Next highest concentration is 2,233 dpm/m3. 

j. "No personnel exposure" to highest/next highest concentrations of 93,887 and 45,997 dpm/m3

. Next highest 

concentration is 21,553 dpm/m3

. 

ATTACHMENT B

AIR CONCENTRATIONS IN SELECTED BUILDINGS

Page 2 of 5 

ATTACHMENT B

AIR CONCENTRATIONS IN SELECTED BUILDINGS

Page 3 of 5 
Document No. ORAUT-TKBS-0010 5 Revision No. 01 Effective Date: 10/15/2009 Page 117 of 150

Table B-2. MACs for selected areas (dpm/m3).

Year

TA-1 (dpm/m3) TA-3 (dpm/m3) TA-21 (dpm/m3)

U Bldg 

U Labs

D Bldg 

EU Labs

D Bldg 

Pu/U Labs

TU Bldg 

normal U

Sigma Bldg 

U/EU Labs

M Bldg 

EU Labs

CMR

U and EU

CMR 

Pu Labs and Waste 

Treatment/ Laundry

DP West 

EU

DP West 

Pu and 

Am CMR-4

DP East

U Bldg 54

DP East

Po

1945 70a 1,500 Po

1947 70a 400

1950 70a 400 400 400

8/1950 68 68 68 68

9/1950 70; Pu, Am, Po, U 68 70 68 1,400 Po

4/1951 8.8a 8 66 66 66 8.8a 8.8a 8.8a 1,400b 1,400 Po

5/1951 8.8 8.8 8.8

10/1951 66 66 8.8

4/1952 66 U

8.8 Pu

66 8.8

1961 66 U-235c

1963 9.0

1964 66 (AL) 

1967 4.0

a. Based on the tolerance limit for plutonium from AEC Division of Biology and Medicine (Shipman et al. 1951).

b. Based on 0.7 cpm/L MAC for room air (Shipman et al. 1951).

c. Wing 9.

ATTACHMENT B

AIR CONCENTRATIONS IN SELECTED BUILDINGS

Page 4 of 5 
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 118 of 150

Table B-3. MACs and ALs for 1964 (dpm/m3

) (Meyer 1964).

Nuclide AL DP West CMR Sigma (TA-3) Shop 13

NU (sol) 66 154.0 154.0 154.0

NU (insol) 66 132.0 132.0 132.0

EU (sol) 66 1,100.0

EU (insol) 66 220.0

U-233 (sol) 66 1,100.0

U-233 (insol) 66 220.0

I-131 NA 2.0E04

Fission products NA 6.7E03

Am-241 (sol) 8.8 13.2

Am-241 (insol) 8.8 220.0

Pu-239 (sol) 8.8 4.4

Pu-239 (insol) 8.8 88.0

Cm-244 (sol) 8.8 19.8

Cm-244 (insol) 8.8 220.0

In 1948, the tolerance level for beta-gamma emitters in air was 1 × 10-7 μCi/cm3 (2.22 × 105 dpm/m3). 

In December 1961, the MAC for MFPs was established in CMR Building, Wing 9 at 6.7 × 103 dpm/m3

. 

In January 1964, the MAC (DP West, Room 401) for 131I was 2.9 × 104 dpm/m3. In January 1968, a 

MAC for 131I was established at 6.7 × 103 dpm/m3

. 

ATTACHMENT B

AIR CONCENTRATIONS IN SELECTED BUILDINGS

Page 5 of 5 
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 119 of 150

Fission and Activation Product Analysis

The first gross beta urine count was devised in 1947 (LASL 1947). Urinalysis was the method used to 

monitor for intakes of fission products until whole-body counting was fully implemented in 1970. A 

procedure dated 1958 lists gross beta urine analysis from 90Sr/Y, 140Ba/La, 144Ce/Pr, 89Sr, and gross 

fission products (McClelland 1958). This might be the same as the procedure referenced in 1947. No 

sensitivity is listed for this procedure. According to the procedure, background was counted before 

and after each sample. Healy (1970) lists nonspecific sensitivities of 50 to 100 dpm/L with an 

investigation level of greater than 200 dpm/L. A similar procedure of oxalate co-precipitation and beta 

counting, effective in 1974 (Gautier 1983), lists a sensitivity of 1 to 2 dps/L and an MDA of 25 pCi/L. 

This procedure might have been effective as early as 1950. No further mention of a gross fission 

product procedure has been found. However, a procedure for 137Cs in urine using gamma 

spectroscopy lists an effective date of May 1965 and an MDA of 100 pCi/L (Gautier 1983).

Reactors operated from 1944 to 1992. Records of stack releases exist for 131I and other fission 

products. Room air concentrations and urine bioassay results have been found in office memoranda 

(LASL 1959). Fission product air contamination around the Omega Water Boiler was noted as a 

“special problem” because ordinary filter paper failed to pick up the gas in the air. This air 

contamination was thought to have occurred because of residual 235U from an old spill that was 

lodged in the interstices of the reactor (Shipman et al. 1951).

Some individuals might have been exposed to fission products while off the Laboratory site or during 

weapons testing. In addition, a potential for exposure to 140Ba/La and 90Sr/Y exists for those 

participating in the RaLa chemistry and testing, 1944 through 1963, primarily at TA-10.

Urine bioassay samples were submitted for gross beta and gamma counting. Samples were analyzed 

by gamma spectroscopy (MDAs for specific geometry and nuclides are unknown.) The sample was 

then processed through alkaline phosphate precipitation and counted on a gas flow counter with a 

background of 530 cpm (40% efficiency). This would be an MDA of approximately 120 dpm/sample. 

Samples were counted at least five times over a 3-week interval to determine the half-life of the 

nuclides. For example, if the sample exhibited a 12-day half-life, 140Ba was recorded; for an 8-day 

half-life, 131I was recorded. Results had a background of 40K subtracted (typically 20 cpm/1,500 ml) 

(LASL 1959).

Gas leaks were detected with filter papers and charcoal, which were analyzed by gamma 

spectroscopy. Personnel potentially exposed were analyzed in the human spectrometer (HUMCO). 

Individuals showing elevated background in the HUMCO were counted in the NaI(Tl) counter [9.5-in. 

by 6-in. NaI(Tl) crystal]. Iodine-135 (6.57-hour half-life)/135Xe (9-hour half-life), 131I, and 133I were 

observed in individuals exposed to gas leaks at reactors (LASL 1959; Van Dilla 1959). Through the 

late 1940s and as late as 1961, before the use of charcoal cartridges, iodine was analyzed from the 

paper filter with an assumed collection efficiency of 0.1%. Radioiodine and noble gases are released 

from facilities that performed fission product chemistry [Wing 9, CMR (TA-3) and TA-48], medical 

isotope preparation (TA-53), and research reactors (TA-35). There is a very small 131I release 

(environmental release of less than 1 mCi/yr) from TA-48 operations, and the Omega West Reactor 

(OWR) Facility (TA-2) is a source of radioiodine.

ATTACHMENT C

BIOASSAY FOR RADIONUCLIDES OTHER THAN PRIMARY RADIONUCLIDES

Page 1 of 5 
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 120 of 150

Interpretation of the fission and activation product urinalysis in a way that is meaningful, as 

representative of all the possible fission and activation products to which a worker might theoretically 

have been exposed, is a challenge. The gross beta procedure separated and counted radionuclides 

of alkaline earths and rare earths, such as strontium, yttrium, barium, lanthanum, and cerium. The 

procedure did not account for the radionuclides of ruthenium, cesium, zinc, cobalt, manganese, 

niobium, or zirconium. The abundances of all the fission products, in relation to each other, varied 

considerably. Certain reactors at LANL operated only briefly. Some exposure might have occurred 

during decommissioning operations and operations with weapons testing. No discussion is available 

for the interpretation of fission product mixtures. However, interviews with current and past LANL 

personnel involved with bioassay indicate that fission products were not considered a significant 

source term for intake among LANL workers.

During the late 1940s and early 1950s, atmospheric testing of nuclear weapons was still being done. 

After such tests in Nevada in the 1950s, these radionuclides were found in urine bioassays for gross 

fission products. The background levels, which were variable, provide a complicating factor for the 

use of gross fission product analyses for dose reconstruction. No specific guidance is available from 

LANL on nuclides or amounts to subtract from bioassay samples. Therefore, any activity detected in 

gross beta urine counts will have to be considered occupational.

The HUMCO whole-body counter was used for screening beginning in the 1950s with follow-up by a 

4-in. by 8–in. NaI(Tl) detector. Body counting and thyroid counting in the current form were not 

available until 1970. Once whole-body counting was established, fission and activation products were 

more often evaluated by whole-body counting. If fission and activation product potential exposure is 

indicated, refer to Section 5.4 for in vivo bioassay sensitivities. The MDAs are listed for many of the 

fission and activation product nuclides (Vasilik and Aikin 1983).

At present, the Bioassay Repository database does not contain any MFP urine bioassay results, and 

there is no plan to supply results for workers unless specific requests are made for the results. 

Unless the telephone interview indicates that the worker might have received bioassay for MFPs, 

there is no way of knowing if these results exist for the worker. After the mid-1950s, when the 

HUMCO and the NaI(Tl) detectors were first placed in service, in vivo bioassay might have been used 

to assess the potential of intakes. The results of any in vivo bioassay performed before 1969 are not 

readily available (Hoover 2007).

In addition to workers at the reactor areas, assessment of internal dose or potential internal dose for 

workers involved in exposure to ionizing radiation associated with RaLa operations at TA-10 (Bayo 

Canyon Site), TA-35 (Ten Site), and Buildings H, Sigma, and U (in TA- 1) is not possible due to the 

lack of urine bioassay monitoring between September 1, 1944, and July 18, 1963, and the 

unavailability of in vivo monitoring until the mid-1950s (NIOSH 2006). These workers, primarily 

chemists, but including other operations and maintenance workers and security personnel, had the 

potential for exposure to 90Sr, 89Sr, 140Ba, and 140La.

Similarly, sufficient records for radionuclides other than tritium, polonium, plutonium, and uranium 

have not been located that would allow the reconstruction of internal doses for most TAs with a history 

of radioactive material use. For this reason, only partial dose reconstructions are possible for LANL

workers with employment before January 1, 1976 (see section 5.1.3).

ATTACHMENT C

BIOASSAY FOR RADIONUCLIDES OTHER THAN PRIMARY RADIONUCLIDES

Page 2 of 5 
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 121 of 150

Strontium

Records of 90Sr urinalyses, routine or special, are very sparse. Inkret et al. (1998c) stated that four 

bioassays were performed for 90Sr in 1997. It is estimated that approximately 200 bioassays were 

performed for 90Sr during the history of the site. The historical compilations of procedures do not list a 

specific 90Sr urinalysis procedure. It appears that any record of 90Sr analysis indicates that LANL 

performed a gross beta analysis (which was determined to be representative of 90Sr) or sent a sample 

to an outside laboratory. Strontium-90 dose currently can be reconstructed only when 90Sr results are 

listed for an individual. No information is available on an MDA for 90Sr analysis. Section 5.4 

discusses the early HUMCO whole-body counters as being capable of counting the bremsstrahlung 

from 90Sr.

Exposure to 90Sr can be expected for persons working in the RaLa Program because 90Sr 

contamination was present in the material used for the program. Persons frequenting the area of the 

shots and those involved with the extraction chemistry would have had a potential for exposure to and 

intake of 90Sr. Persons from TA-1, Sigma, H, and U Buildings, where RaLa operations were 

conducted, would also have the potential for intakes of 140Ba and 140La. The RaLa sources were 

prepared by Group G-7 or Group G-6 workers at the TA-10 Chemical Process Building from 1944 to 

1950. This function moved to TA-35 (Ten Site) from 1951 to 1963. The name of the site is likely tied 

to the operating group, CMR-10. The CMR-10 group relocated to Ten Site sometime between April 

and December 1950. The Idaho Chemical Processing Plant became the source of purified 140Ba in 

1956, and a typical shipment was about 40,000 Ci of 140Ba. The 140La sources prepared at Ten Site 

were usually in the range of 2,000 to 4,000 Ci. Almost 2 million Ci of 140Ba had been handled at Ten 

Site by 1963 when the RaLa Program was terminated. The TA-35 RaLa cell and control room have 

been completely dismantled. Barium-140 and 140La would have been detectable in the HUMCO 

screening and quantified in the NaI follow-up, although no MDA is available for those nuclides. The 

Savannah River Site lists an MDA of 9.3 nCi for that period (ORAUT 2005c). However, it is not 

possible to reconstruct internal dose or potential internal dose from 90Sr for workers with potential 

exposure to ionizing radiation associated with RaLa operations at TA-10 (Bayo Canyon Site), TA-35 

(Ten Site), and Buildings H, Sigma, and U (in TA-1)

Actinium

Notebook 6489, containing 227Ac in vitro bioassay results, has been found. Only 15 analyses are 

recorded in the notebook. All analyses were performed in 1954. No other records of 227Ac bioassay

have been found. Results listed were 0 to 0.6 dpm/sample.

Potential worker encounters with 227Ac occurred during the decommissioning of TA-21-153 in 1978. 

However, all workers on the decommissioning wore full-face particulate respirators and 

anticontamination clothing. Workers were regularly monitored with nasal swipes and in vivo bioassay 

during the project. Results of the bioassay and nasal swipes were below detection levels (Harper and 

Garde 1981).

Shipman et al. (1951) mentions the determination of 226Ra as the only practical method “at present” 

for estimating individual exposures to actinium.

ATTACHMENT C

BIOASSAY FOR RADIONUCLIDES OTHER THAN PRIMARY RADIONUCLIDES

Page 3 of 5 
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 122 of 150

No other information is available. Results for any bioassay that might have been performed for the 

individual will not be present in the Bioassay Repository database. These results will be available 

only by special request.

Phosphorus-32 and Carbon-14

Both of these radionuclides were encountered as labeled compounds. Standard biokinetics do not 

apply to labeled compounds. A bioassay procedure for 32P in urine lists an effective date of August 

1975 and an MDA of 40 pCi/L. This is also the MDA listed in a procedure in 1983 (Gautier 1983). No 

bioassay procedure is found for 14C.

Results for any bioassay that might have been performed for the individual will not be present in the 

Bioassay Repository database. These results will be available only by special request.

Thorium

A procedure for a coprecipitation technique and a colorimetric final determination was listed as 

effective January 1963. The MDA for this analysis was listed as 20 μg/L. Dummer (1958) listed the 

MDA for the colorimetric procedure as 0.01 μg/L. While plutonium can coprecipitate with thorium, the 

colorimetric determination prevented any interference from the plutonium. An earlier procedure titled 

“The Determination of Thorium230 in Urine” was listed in McClelland (1958). This procedure used low 

background alpha proportional counting and lists a sensitivity of 0.05 dpm (no volume units) of 230Th. 

The procedure stated that natural thorium, plutonium, some americium, curium, actinium, and 

neptunium are carried over with the precipitate.

Only the results of the radiometric analysis of 44 urine samples for 230Th performed in 1958 have been 

found thus far. A sensitivity of 0.2 dpm (no volume units) was listed for these analyses. Dummer 

(1958) suggested the use of alpha proportional counting for 230Th bioassay samples. Thorium-230 is 

also referred to as ionium. Results for any bioassay that might have been performed for the individual 

will not be present in the Bioassay Repository database. These results are available only by special 

request.

Shipman et al. (1951) discussed monitoring of thorium daughters in air. Daughter products would be 

produced from the decay of 232Th. Casting, machining, and other operations with thorium metal were 

undertaken in 1951. This suggests that 232Th was in use during this period. Thorium-232 has been 

identified in environmental emissions from waste sites at LANL. After 1980, some work with thoriumbased materials was performed. The thorium was probably 232Th.

The isotopic mixture should be based on the facility. Conversations with former LANL personnel 

suggest that most exposure to thorium would be incidental to the exposure to decay products of 

uranium. However, statements about work with thorium metal would seem to indicate otherwise. 

Building 159, Sigma Complex, is designated as a Thorium Storage Building for the storage of ingot 

and oxide forms, which would suggest 232Th and daughters. Therefore, the dose reconstructor should 

use best judgment for selection of the isotope or mixture. Thorium-230 is the maximizing conversion 

from mass units. However, it is unlikely, given the current structure of the database, that results from 

thorium bioassays will ever be submitted for a worker unless specific requests are made for the 

results. In that situation, the isotope that was encountered would probably be known.

ATTACHMENT C

BIOASSAY FOR RADIONUCLIDES OTHER THAN PRIMARY RADIONUCLIDES

Page 4 of 5 
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 123 of 150

The Tiger Team Assessment report (DOE 1991) indicated that line managers were not aware that 

thorium and its decay products were internal radiological hazards and that workers handling gram 

quantities of dispersible thorium oxide powders and other thorium compounds should be identified by 

the checklist system for participation in the bioassay program. 

Default absorption types for thorium are M and S. The absorption type should be selected based on 

the compound expected or the matrix (in the case of 230Th incidental to uranium exposure).

If radiometric alpha spectroscopy analysis, rather than total alpha, results of 232Th are encountered, 

the contribution from decay product nuclides should be included. Thorium-232 decays with a 6.7-year 

half-life to 228Ra, which decays with a 5.75-year half-life to 6.13-hour 228Ac, which decays to 1.91-year 

228Th. See C-1.

Table C-1. Specific activity of thorium isotopes. 

Isotope Specific activity (pCi/μg)

Th-228 8.1946E+08

Th-230 2.0184E+04

Th-232 1.0966E-01

Protactinium-231

A procedure for a coprecipitation technique and proportional alpha counting is documented in 

McClelland (1958). The MDA is listed as 0.88 dpm/24-hr sample. No other mention of this analysis is 

found in later procedure documents or reports.

Results for any bioassay that might have been performed for the individual will not be present in the 

Bioassay Repository database. These results are available only by special request.

ATTACHMENT C

BIOASSAY FOR RADIONUCLIDES OTHER THAN PRIMARY RADIONUCLIDES

Page 5 of 5 
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 124 of 150

ATTACHMENT D

RESPIRATORY PROTECTION PROGRAM

Page 1 of 2

Respiratory Protection Program

The first reference to the need for respiratory protection for LANL workers was in a memorandum

requesting “assault-type gas masks for testing in connection with toxic effects of X10 (plutonium) dust” 

(Oppenheimer 1944). The May 31, 1944, “Health Safety Rules” for Buildings H and V-8 specified that 

“[a]n approved respirator or gas mask must be worn during operations requiring the exposure of the 

dry material to the air and during wet operations involving spray or splatter” (LASL 1944a). Positive 

pressure masks were being introduced into the program at that time (LASL 1944a). Dry boxes and 

respirators were a satisfactory method of handling dry material; however, individuals who worked with 

solutions consistently received the highest exposure despite all precautions.

Respiratory protective devices were provided during the Manhattan Project by the U.S. Army 

Chemical Warfare Laboratories. Use of the M-9 mask with either the M-11 canister for gases and 

particulates or the M-14 canister for particulates began in 1946. The efficiency of the canister was 1 

in 100,000 particles of dioctyl phthalate (Adley et al. 1963). This mask continued to be used as a 

standard until commercially produced masks were available. By the time the handbook on radiation 

protection (Dummer 1958) was published, a variety of U.S. Bureau of Mines-approved commercial 

respirators and self-contained breathing apparatus was available. Approved respirators were used 

where possible. The Industrial Hygiene group maintained facilities for testing respiratory protection 

equipment and could make recommendations on equipment. The use of Wilson-type respirators was 

discussed in 1973 (Hempelmann, Richmond, and Voelz 1973).

The Industrial Hygiene group at LANL continued to be at the forefront of development of respiratory 

devices and programs. Respiratory protection programs are documented beginning with Dummer 

(1958). The programs included training in the use of and fit testing for air-purifying respirators and 

self-contained breathing apparatus. Protection factors assumed for respiratory protection equipment 

are listed in Table D-1. The protection factors appear to remain consistent throughout the history of 

the program within the exceptions noted.

Table D-1. Respiratory protection apparatus

Equipment Protection factor Remarks

Half-mask Not more than 10 × MPCa b, c

Full-face mask Not more than 50 × MPC (Dummer 1958)d b

Full-face mask Not more than 100 ×, MPC (Healy 1970; LANL 1984)e b

Full-face mask with fit field tested Not more than 1,000 × MPC (Healy 1970) b

Full-face mask with chemical canister Used for organic gases and vapors up to 2% b

Air-line equipment Used above limits of air purifying respirators. No specific 

protection factors listed

Self-contained breathing apparatus Used above limits of air purifying respirators. No specific 

protection factors listed

a. Assumed 20% penetration because of fitting limitations and 50% workday usage.

b. Air purifying respirators shall never be used in an atmosphere immediately dangerous to life and health.

c. Half-face respirators were not permitted to be used in airborne highly toxic dust such as plutonium (Shipman 1964). 

LANL (2004a) stated that half-face respirators are not approved for use with radiological materials or airborne 

radioactivity.

d. Only one style was available at that time; masks were assumed to be only 98% to 99% efficient.

e. Powered air-purifying respirators not more than 3,000 times with high-efficiency particulate air (HEPA) filters.
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 125 of 150

Respirator mask fit and proper usage remained a significant challenge throughout the development of 

the program. In addition, the decision to don a respirator is often based on the alarm of a continuous 

air monitor in the area.

ATTACHMENT D

RESPIRATORY PROTECTION PROGRAM

Page 2 of 2 
Document No. ORAUT-TKBS-0010 5 Revision No. 01 Effective Date: 10/15/2009 Page 126 of 150

Each item in the tables below is a summary of a programmatic problem or observation made by the ‘Tiger Team’ assessment described in 

DOE (1991). The dose reconstructor should evaluate each claim to determine whether the programmatic findings listed may imply an 

unmonitored intake in accordance with Section 5.5 of this technical basis document. For convenience, those findings with potential to 

affect dose reconstructions are listed in Table E-1; all problems or observations are grouped by topic in Tables E-2 through E-9 (those 

requiring evaluation are thus repeated in Tables E-2 through E-9).

NOTE: The following acronyms are used to save space in the table:

NSA means no specific action;

PUI means potential unmonitored intake.

Table E-1. Summary Table: All Programmatic items requiring dose reconstructor actions. 

Item Finding/issue description

Dose 

affected DR action Comments

1-1 ...The potential for monitored and unmonitored intakes has 

existed throughout the history of the site. 

All Each 

claim is 

evaluated 

for PUI.

1-2 Only workers with a significant potential for exposure were monitored. 

Although the number of individuals monitored has increased, not all 

individuals working at LANL are currently monitored. 

All (1988 

and after)

Each 

claim is 

evaluated 

for PUI.

Like all facilities, LANL prioritizes 

internal dose monitoring for those 

individuals most likely to be exposed. 

1-3 Other incident information might not have been captured in the 

database.

All Each 

claim is 

evaluated 

for PUI.

1-4 Before the 1970s, individuals were assigned to a bioassay program as 

determined by the area health physics monitors. ... instances might 

have occurred, especially in the early history, in which a person not 

normally assigned to radiation work was asked to participate as a 

substitute in a task involving radiation or radioactive materials. These 

persons were not likely to have regularly, or possibly ever, participated 

in the bioassay program. It is possible that their participation in these 

tasks was never recorded. 

All Each 

claim is 

evaluated 

for PUI.

ATTACHMENT E

PROGRAMMATIC ISSUES TO BE REVIEWED FOR EVALUATING FOR POTENTIAL UNMONITORED INTAKES

Page 1 of 19
Document No. ORAUT-TKBS-0010 5 Revision No. 01 Effective Date: 10/15/2009 Page 127 of 150

Item Finding/issue description

Dose 

affected DR action Comments

1-5 Short-term workers, such as summer students, persons participating in 

postdoctoral work, contractors, and teachers, might not have fully 

participated in routine bioassay programs. These workers were 

monitored for internal exposure only in unusual circumstances. Near 

his/her termination date, the worker might have received an in vivo

count. The worker might have been required to submit an initial urine 

sample or have an initial in vivo count. 

All Each 

claim is 

evaluated 

for PUI.

1-6 The report also indicated that line managers did not make appropriate 

changes in the staff member dosimetry requirements using the 

Employee Health Physics Checklist, as required, in the event of 

intragroup transfers or job content changes (DOE 1991). 

All Each 

claim is 

evaluated 

for PUI.

1-7 Radiation protection technicians were not reviewing operations that 

involved radioactive materials to ensure that workers who should 

participate in the bioassay program did so. 

All Each 

claim is 

evaluated 

for PUI.

1-8 Low priority was given to the enforcement of the requirement.

In 1991, DOE issued a report after the 1990 Tiger Team Assessment of 

the Laboratory. 

All Each 

claim is 

evaluated 

for PUI.

1-9 The contamination control programs at the sites did not ensure 

complete control of the spread of contamination. Several instances 

were cited (DOE 1991) in which barriers had been removed or signage 

was inappropriate or missing. This included reactor and tritium 

facilities. These instances would allow general personnel access to 

controlled or contaminated areas and allow worker egress from these 

areas and ingress to uncontrolled areas without appropriate monitoring. 

All Each 

claim is 

evaluated 

for PUI.

1-10 In addition, instances in which tools used in maintenance work that 

involved radioactive material contamination are not uniquely identified 

to prevent use in an uncontrolled area.

All Each 

claim is 

evaluated 

for PUI.

Workplace monitoring not used in dose 

reconstruction; release of contaminated 

material could result in small intakes. 

ATTACHMENT E

PROGRAMMATIC ISSUES TO BE REVIEWED FOR EVALUATING FOR POTENTIAL UNMONITORED INTAKES

Page 2 of 19
Document No. ORAUT-TKBS-0010 5 Revision No. 01 Effective Date: 10/15/2009 Page 128 of 150

Item Finding/issue description

Dose 

affected DR action Comments

1-11 Posting was inconsistent throughout LANL, and line management failed 

to provide the necessary resources to ensure that the posting 

requirements were uniformly implemented in the zoning and posting of 

each facility (DOE 1991). 

All Each 

claim is 

evaluated 

for PUI.

Workplace monitoring not used in dose 

reconstruction; release of contaminated 

material could result in small intakes. 

1-12 The frequency of surveys mandated by procedures was not being 

consistently followed. Semiannual floor surveys required by procedure 

had not been performed in some areas in 1991. 

All Each 

claim is 

evaluated 

for PUI.

Workplace monitoring not used in dose 

reconstruction; release of contaminated 

material could result in small intakes. 

1-13 The assessment (DOE 1991) indicated that signs and labels throughout 

the plutonium and uranium facilities did not indicate radiological 

conditions and were not accurate or had other problems that could lead 

to unsafe practices. Examples were a storage area posted as 

“Chemical Waste Only” that had some containers with radiation labels 

requiring alpha monitoring. Radiation Area signs on individual doors 

did not have radiological conditions posted on them. Localized 

contamination areas – radiological hot spots – were not labeled with 

radiation-type markings.

All Each 

claim is 

evaluated 

for PUI.

Workplace monitoring not used in dose 

reconstruction; release of contaminated 

material could result in small intakes. 

1-14 The documentation of smear surveys was not consistently performed 

(DOE 1991), including high count (greater than 50,000 cpm) areas in 

PF-4, which was used as a staging area for radioactive waste and 

contaminated equipment. 

All Each 

claim is 

evaluated 

for PUI.

Workplace monitoring not used in dose 

reconstruction; release of contaminated 

material could result in small intakes. 

1-15 In addition, instances might have occurred, especially in the early 

history, in which a person not normally assigned to radiation work was 

asked to participate as a substitute in a task that involved radiation or 

radioactive materials. These workers were not likely to have regularly, 

or ever, participated in the bioassay program or potentially have their 

participation in the task recorded. Indications of this type of exposure 

might come from claimant interviews

All Each 

claim is 

evaluated 

for PUI.

ATTACHMENT E

PROGRAMMATIC ISSUES TO BE REVIEWED FOR EVALUATING FOR POTENTIAL UNMONITORED INTAKES

Page 3 of 19
Document No. ORAUT-TKBS-0010 5 Revision No. 01 Effective Date: 10/15/2009 Page 129 of 150

Item Finding/issue description

Dose 

affected DR action Comments

1-16 The formal protocol for placing workers in bioassay programs was in 

place in 1970 with the Health Physics Checklist. Before the Health 

Physics Checklist, a person was placed in a bioassay program at the 

discretion of the health physics monitor in charge of the area. While 

these protocols ensured that most workers who required monitoring 

were monitored, some workers might not have been included. In 

addition, there is a possibility that the checklist might not have been 

updated when a worker’s job title or responsibilities changed (DOE 

1991). 

All Each 

claim is 

evaluated 

for PUI.

1-17 Therefore, exposure to some amount of airborne contamination is a 

reasonable assumption for all workers during this early era, regardless 

of job title.

Alpha 

emitters 

1943-45

Evaluate 

claim for 

PUI

Applicable only to partial dose 

reconstructions during SEC period.

1-18 The Tiger Team Assessment (DOE 1991) indicated that while in some 

areas personnel were either evaluated for participation or all personnel 

participated in appropriate bioassay programs (TA-55 and TA-50), not 

all personnel at the plutonium and DU facilities were evaluated for 

participation in bioassay programs. 

Pu and U Evaluate 

claim for 

PUI

1-19 The potential for unmonitored intakes was significant in the early years 

(1944 to 1946) for any site worker. Bioassay was provided only for 

those workers who were most exposed. However, because the 

excretion of plutonium is continuous after an intake, significant intakes 

might have been identified on later routine bioassay samples with 

improved sensitivity for workers who remained at LANL. 

Pu and U Evaluate 

claim for 

PUI

1-20 As bioassay sensitivities and respiratory protection equipment 

improved, the potential for intakes decreased. Due to rigorous 

workplace monitoring, the probability that a worker received a large 

intake of radioactive material that was unmonitored and unnoticed was 

less after 1946. However, the probability of unmonitored small intakes 

was larger. Periodic reports from H Division of air samples, 

contamination incidents, and hot spots continued to identify a 

significant number of “over tolerance” occurrences throughout the 

1960s. 

Pu and U Evaluate 

claim for 

PUI

ATTACHMENT E

PROGRAMMATIC ISSUES TO BE REVIEWED FOR EVALUATING FOR POTENTIAL UNMONITORED INTAKES

Page 4 of 19
Document No. ORAUT-TKBS-0010 5 Revision No. 01 Effective Date: 10/15/2009 Page 130 of 150

Item Finding/issue description

Dose 

affected DR action Comments

1-21 Review of Summary of Radiological Incident Reports (RIR) through 

June 1998 indicates that, while the number of RIRs had decreased 

significantly, the potential for unmonitored small intakes continues to 

exist (Bates 1998).

Pu and U Evaluate 

claim for 

PUI

1-22 The Tiger Team Assessment report (DOE 1991) indicated that line 

managers were not aware that thorium and its decay products were 

internal radiological hazards and that workers handling gram quantities 

of dispersible thorium oxide powders and other thorium compounds 

should be identified by the checklist system for participation in the 

bioassay program. 

Thorium Apply 

thorium 

doses 

when 

indicated

Thorium is specifically listed as a 

potential unmonitored actinide source, 

and assignment of a maximum dose is 

described in Section 5.5 of the TBD. 

Table E-2 Programmatic items potentially affecting the calculation or assignment of dose from actinides. 

Item Finding/issue description

Dose 

affected DR action Comments

2-1 In 1951, an internal memorandum was issued requesting cooperation 

with Health Rules (Tribby 1946a) that had apparently been violated 

(Vier 1951). Included in the areas where compliance was requested 

were:

• Getting nose counts every day if you are in the lab at the 

time except on rare occasions when it would seriously 

interfere with your work.

• Submitting urine samples at least once a month.

• Checking hand counts and/or washing hands before 

leaving Bldg. 52. 

Actinides in 

TA-21-152

None NIOSH has determined that it is not 

feasible to reconstruct dose from 

actinides in the pre-1976 timeframe. 

2-2 In TA-48, Building 1, Radiochemistry, alpha hood ventilation was 

controlled in a reactive mode by the changing of HEPA filters based on 

periodic airflow checks; the system had no real-time equipment with

alarms to notify occupants if ventilation was inadequate.

Actinides NSA Higher levels of airborne radioactivity 

could result; but no circumstance is 

described that indicates whether actual 

exposures resulted. Doses based on 

bioassay would reflect higher intakes. 

ATTACHMENT E

PROGRAMMATIC ISSUES TO BE REVIEWED FOR EVALUATING FOR POTENTIAL UNMONITORED INTAKES

Page 5 of 19
Document No. ORAUT-TKBS-0010 5 Revision No. 01 Effective Date: 10/15/2009 Page 131 of 150

Table E-3. Programmatic items potentially affecting the calculation or assignment of dose from all radionuclides. 

Item Finding/issue description

Dose 

affected DR action Comments

3-1 Early Health Group reports indicate contamination inside many 

respirators, indicating improper storage and handling and poor fit during 

use (LASL 1944b)

All None No credit is taken in dose 

reconstruction for respiratory 

protection.

3-2 Nevertheless, the potential for monitored and unmonitored intakes has 

existed throughout the history of the site. 

All Each 

claim is 

evaluated 

for PUI.

3-3 Only workers with a significant potential for exposure were monitored. 

Although the number of individuals monitored has increased, not all 

individuals working at LANL are currently monitored. 

All (1988 

and after)

Each 

claim is 

evaluated 

for PUI.

Like all facilities, LANL prioritizes 

internal dose monitoring for those 

individuals most likely to be exposed. 

3-4 Other incident information might not have been captured in the 

database.

All Each 

claim is 

evaluated 

for PUI.

3-5 Before the 1970s, individuals were assigned to a bioassay program as 

determined by the area health physics monitors. ... instances might 

have occurred, especially in the early history, in which a person not 

normally assigned to radiation work was asked to participate as a 

substitute in a task involving radiation or radioactive materials. These 

persons were not likely to have regularly, or possibly ever, participated 

in the bioassay program. It is possible that their participation in these 

tasks was never recorded. 

All Each 

claim is 

evaluated 

for PUI.

ATTACHMENT E

PROGRAMMATIC ISSUES TO BE REVIEWED FOR EVALUATING FOR POTENTIAL UNMONITORED INTAKES

Page 6 of 19
Document No. ORAUT-TKBS-0010 5 Revision No. 01 Effective Date: 10/15/2009 Page 132 of 150

Item Finding/issue description

Dose 

affected DR action Comments

3-6 Short-term workers, such as summer students, persons participating in 

postdoctoral work, contractors, and teachers, might not have fully 

participated in routine bioassay programs. These workers were 

monitored for internal exposure only in unusual circumstances. Near 

his/her termination date, the worker might have received an in vivo

count. The worker might have been required to submit an initial urine 

sample or have an initial in vivo count. 

All Each 

claim is

evaluated 

for PUI.

3-7 The report also indicated that line managers did not make appropriate 

changes in the staff member dosimetry requirements using the 

Employee Health Physics Checklist, as required, in the event of 

intragroup transfers or job content changes (DOE 1991). 

All Each 

claim is 

evaluated 

for PUI.

3-8 Radiation protection technicians were not reviewing operations that 

involved radioactive materials to ensure that workers who should 

participate in the bioassay program did so. 

All Each 

claim is 

evaluated 

for PUI.

3-9 Low priority was given to the enforcement of the requirement.

In 1991, DOE issued a report after the 1990 Tiger Team Assessment of 

the Laboratory. 

All Each 

claim is 

evaluated 

for PUI.

3-10 The contamination control programs at the sites did not ensure 

complete control of the spread of contamination. Several instances 

were cited (DOE 1991) in which barriers had been removed or signage 

was inappropriate or missing. This included reactor and tritium 

facilities. These instances would allow general personnel access to 

controlled or contaminated areas and allow worker egress from these 

areas and ingress to uncontrolled areas without appropriate monitoring. 

All Each 

claim is 

evaluated 

for PUI.

3-11 In addition, instances in which tools used in maintenance work that 

involved radioactive material contamination are not uniquely identified 

to prevent use in an uncontrolled area.

All Each 

claim is 

evaluated 

for PUI.

Workplace monitoring not used in dose 

reconstruction; release of contaminated 

material could result in small intakes. 

ATTACHMENT E

PROGRAMMATIC ISSUES TO BE REVIEWED FOR EVALUATING FOR POTENTIAL UNMONITORED INTAKES

Page 7 of 19
Document No. ORAUT-TKBS-0010 5 Revision No. 01 Effective Date: 10/15/2009 Page 133 of 150

Item Finding/issue description

Dose 

affected DR action Comments

3-12 Posting was inconsistent throughout LANL, and line management failed 

to provide the necessary resources to ensure that the posting 

requirements were uniformly implemented in the zoning and posting of 

each facility (DOE 1991). 

All Each 

claim is 

evaluated 

for PUI.

Workplace monitoring not used in dose 

reconstruction; release of contaminated 

material could result in small intakes. 

3-13 The frequency of surveys mandated by procedures was not being 

consistently followed. Semiannual floor surveys required by procedure 

had not been performed in some areas in 1991. 

All Each 

claim is 

evaluated 

for PUI.

Workplace monitoring not used in dose 

reconstruction; release of contaminated 

material could result in small intakes. 

3-14 The assessment (DOE 1991) indicated that signs and labels throughout 

the plutonium and uranium facilities did not indicate radiological 

conditions and were not accurate or had other problems that could lead 

to unsafe practices. Examples were a storage area posted as 

“Chemical Waste Only” that had some containers with radiation labels 

requiring alpha monitoring. Radiation Area signs on individual doors 

did not have radiological conditions posted on them. Localized 

contamination areas – radiological hot spots – were not labeled with 

radiation-type markings.

All Each 

claim is 

evaluated 

for PUI.

Workplace monitoring not used in dose 

reconstruction; release of contaminated 

material could result in small intakes. 

3-15 The documentation of smear surveys was not consistently performed 

(DOE 1991), including high count (greater than 50,000 cpm) areas in 

PF-4, which was used as a staging area for radioactive waste and

contaminated equipment. 

All Each 

claim is 

evaluated 

for PUI.

Workplace monitoring not used in dose 

reconstruction; release of contaminated 

material could result in small intakes. 

3-16 Airflow to hoods was not verified in the TA-21 Chemistry Laboratory. All NSA The poor practice described may 

increase the amount of dose received, 

but does not affect the estimation of 

occupational internal dose.

3-17 The ion chamber survey instrument used at IBML did not have sufficient 

sensitivity to measure activity at DOE release limits and personnel were 

not formally trained in the use of the instrument and proper 

documentation of surveys (DOE 1991). 

All None Workplace monitoring not used in dose 

reconstruction

ATTACHMENT E

PROGRAMMATIC ISSUES TO BE REVIEWED FOR EVALUATING FOR POTENTIAL UNMONITORED INTAKES

Page 8 of 19
Document No. ORAUT-TKBS-0010 5 Revision No. 01 Effective Date: 10/15/2009 Page 134 of 150

Item Finding/issue description

Dose 

affected DR action Comments

3-18 Procedures for the removal of protective clothing were not posted at 

step-off pads in TA-3 and surveying procedures were not being 

followed (DOE 1991). Personnel leaving a contaminated area at the 

Isotope Production Facility and Area A-East in TA-53 were observed 

performing contamination monitoring while wearing plastic booties and 

rubber gloves. No final survey was performed when protective clothing 

was removed (DOE 1991). 

All None Workplace monitoring not used in dose 

reconstruction 

3-19 The use of open-front hoods in the TA-3 CMR Building led to an 

increased frequency of radioactive material contamination incidents. 

All? NSA The poor practice described may 

increase the amount of dose received, 

but does not affect the estimation of 

occupational internal dose.

3-20 The Tiger Team Assessment (DOE 1991) noted that safe operating 

procedures (SOPs) were not available for certain tasks where there 

was a potential for intakes. These SOPs included changing HEPA 

filters on a portable vacuum cleaner at the Sigma Complex in TA-3. 

All NSA The poor practice described may 

increase the amount of dose received, 

but does not affect the estimation of 

occupational internal dose.

3-21 The assessment also noted that the practice existed for implementation 

of procedures involving radiological work without line management or 

Health Physics approval, which was not in compliance with DOE Order 

5480.11 (DOE 1988). 

All None

3-22 While procedures that are not reviewed are not necessarily inadequate, 

the potential can exist for inconsistent application of guidance and 

insufficient protective measures or monitoring during the performance 

of the operations.

All NSA No inadequacy in the procedure is 

stated. 

3-23 The Tiger Team Assessment indicated that procedures and processes 

used to implement radiation protection programs at accelerator facilities 

are adequate to ensure that the requirements of DOE Order 5480.11 

(DOE 1988) and the LANL ESH Manual were being met (DOE 1991). 

No additional details were given in the assessment to determine if 

potential intakes were involved.

All None No actions by the dose reconstructor 

are possible based on this finding. 

3-24 Out-of-calibration instruments, such as glovebox hand and foot 

monitors and tritium monitors, were not taken out of service.

All NSA No actions by the dose reconstructor 

are possible based on this finding.

ATTACHMENT E

PROGRAMMATIC ISSUES TO BE REVIEWED FOR EVALUATING FOR POTENTIAL UNMONITORED INTAKES

Page 9 of 19
Document No. ORAUT-TKBS-0010 5 Revision No. 01 Effective Date: 10/15/2009 Page 135 of 150

Item Finding/issue description

Dose 

affected DR action Comments

3-25 In TA-55 and several other facilities, placement of air monitoring 

instruments at the DU sites was not based on studies of flow patterns. 

All None While responding to an alarm can 

reduce the dose received, this project 

is concerned only with ensuring that 

the dose an individual received is 

reconstructed with sufficient accuracy 

to result in a correct determination of 

whether the probability of causation 

resulting from that dose exceeds or 

does not exceed 50%.

3-26 Facility air monitor alarm points, used in accordance with the 

requirements of DOE Order 5480.11 (DOE 1988) to warn workers that 

airborne radioactive material contamination levels had exceeded an AL, 

were not always set at a uniform level. The set points varied from 

monitor to monitor even in the same building. 

All None While responding to an alarm can 

reduce the dose received, this project 

is concerned only with ensuring that 

the dose an individual received is 

reconstructed with sufficient accuracy 

to result in a correct determination of 

whether the probability of causation 

resulting from that dose exceeds or 

does not exceed 50%. 

3-27 Training programs had not been established to ensure that routine dose 

rate and contamination surveys were conducted in a consistent 

manner.

All None Workplace monitoring is within the 

hierarchy of information used to 

reconstruct dose; however it has been 

used rarely, or not at all, to date on this 

project. 

3-28 The Tiger Team Assessment (DOE 1991) indicated that airflow, particle 

size, and solubility studies had not been documented at all DU, 

plutonium, and firing sites to verify the proper positioning of the CAMs 

for monitoring airborne activity. 

All None This project assumes default particle 

sizes, and bases solubility upon 

favorable to claimant assumptions. 

Airflow considerations are also not 

used as part of dose reconstruction. 

ATTACHMENT E

PROGRAMMATIC ISSUES TO BE REVIEWED FOR EVALUATING FOR POTENTIAL UNMONITORED INTAKES

Page 10 of 19
Document No. ORAUT-TKBS-0010 5 Revision No. 01 Effective Date: 10/15/2009 Page 136 of 150

Item Finding/issue description

Dose 

affected DR action Comments

3-29 In addition, instances might have occurred, especially in the early 

history, in which a person not normally assigned to radiation work was 

asked to participate as a substitute in a task that involved radiation or 

radioactive materials. These workers were not likely to have regularly, 

or ever, participated in the bioassay program or potentially have their 

participation in the task recorded. Indications of this type of exposure 

might come from claimant interviews

All Each 

claim is 

evaluated 

for PUI.

3-30 As late as 1977, it was reported that releasable and nonreleasable 

(contaminated) salvage materials were accessible to “scavenger 

personnel” who frequented these bins and possibly mixed the types of 

scrap while it was waiting for pickup (University of California 1977d). A 

fence was installed to prevent uncontrolled access when an incident 

resulted in contaminated scrap leaving the area on a truck. 

All None The poor practice described does not 

affect the estimation of occupational 

internal dose. 

3-31 The formal protocol for placing workers in bioassay programs was in 

place in 1970 with the Health Physics Checklist. Before the Health 

Physics Checklist, a person was placed in a bioassay program at the 

discretion of the health physics monitor in charge of the area. While 

these protocols ensured that most workers who required monitoring 

were monitored, some workers might not have been included. In 

addition, there is a possibility that the checklist might not have been 

updated when a worker’s job title or responsibilities changed (DOE 

1991). 

All Each 

claim is 

evaluated 

for PUI.

3-32 CAM alarms have been used historically as an indication that a room air 

concentration had changed and that respiratory protection was 

required. Whicker et al. (1997) found that the selection of the location 

of the CAM was critical to the reliability of the response. The typical 

location of CAMs at the ventilation exhaust point of a room was not 

considered optimal. 

All NSA CAM actions are protective, while the 

DR process seeks only to estimate the 

actual dose received.

ATTACHMENT E

PROGRAMMATIC ISSUES TO BE REVIEWED FOR EVALUATING FOR POTENTIAL UNMONITORED INTAKES

Page 11 of 19
Document No. ORAUT-TKBS-0010 5 Revision No. 01 Effective Date: 10/15/2009 Page 137 of 150

Item Finding/issue description

Dose 

affected DR action Comments

3-33 Results also suggested that if a worker caused the release and was at 

or near the release point, the worker could be exposed for a significant 

period before a radioactive cloud reached the CAM. 

All NSA CAM actions are protective, while the 

DR process seeks only to estimate the 

actual dose received.

3-34 The Tiger Team Assessment (DOE 1991) indicated that airflow, particle 

size, and solubility studies had not been documented at all DU, 

plutonium, and firing sites to verify the proper positioning of the CAMs 

for monitoring airborne activity. These sites included TA-50 and TA-3 

(Sigma Complex).

All None This poor practice may add to the dose 

received, but does not affect estimation 

of the doses received.

3-35 Particle size studies had been initiated at Building PF-4 All NSA NIOSH uses default particle sizes as 

documented in TIB-0060 when this 

information is not known. 

3-36 The Tiger Team Assessment indicated a concern that the testing 

program for HEPA filters did not include periodic testing and 

replacement of filters on contaminated vacuum cleaners and pumps 

used at accelerator facilities, nor did it ensure that the filters were 

certified by an approved DOE facility before installation (DOE 1991).

All 

(particulate)

None This poor practice may add to the dose 

received, but does not affect estimation 

of the doses received. 

3-37 In TA-48, Building 1, Radiochemistry, alpha hood ventilation was 

controlled in a reactive mode by the changing of HEPA filters based on 

periodic airflow checks; the system had no real-time equipment with 

alarms to notify occupants if ventilation was inadequate.

All 

(particulate)

None This poor practice may add to the dose 

received, but does not affect estimation 

of the doses received.

Table E-4. Programmatic items potentially affecting the calculation or assignment of dose from ”alpha emitters.” 

Item Finding/issue description

Dose 

affected DR action Comments

4-1 Therefore, exposure to some amount of airborne contamination is a 

reasonable assumption for all workers during this early era, regardless 

of job title.

Alpha 

emitters 

1943-45

Evaluate 

claim for 

PUI

Applicable only to partial dose 

reconstructions during SEC period.

ATTACHMENT E

PROGRAMMATIC ISSUES TO BE REVIEWED FOR EVALUATING FOR POTENTIAL UNMONITORED INTAKES

Page 12 of 19
Document No. ORAUT-TKBS-0010 5 Revision No. 01 Effective Date: 10/15/2009 Page 138 of 150

Table E-5. Programmatic items potentially affecting the calculation or assignment of dose from unknown radionuclides. 

Item Finding/issue description

Dose 

affected DR action Comments

5-1 In addition, personnel at the firing sites, including TA-36 and TA-15, 

were not evaluated for inclusion in bioassay programs. 

None None Radiological hazards in TA-36 and TA15 are limited to flash radiography; 

internal doses from radio frequency

linear accelerators (used to produce Xrays) are negligible. 

5-2 Table A-15 in Attachment A lists the routine sampling frequencies by 

period. In 1978, sampling frequencies were reevaluated for LANL and 

Zia employees. To attempt to reduce the number of sampled Zia 

employees to 500 per year, supervisors who only performed 

inspections were eliminated from the schedule (University of California

1978a). Special and emergency sample frequencies are listed in 

Tables A-19 to A-21 of Attachment A.

None None Pu bioassay was required for work in 

plutonium areas Lawrence 1976). This 

requirement simply reduced the 

number of qualified individuals to save 

cost

5-3 The Tiger Team Assessment indicated a concern that the testing 

program for HEPA filters did not include periodic testing and 

replacement of filters on contaminated vacuum cleaners and pumps 

used at accelerator facilities, nor did it ensure that the filters were 

certified by an approved DOE facility before installation (DOE 1991).

Unknown None This may have resulted in higher levels 

of airborne radioactivity; however, no 

circumstances are described that 

would indicate whether actual 

exposures resulted or did not. Doses 

based on bioassay data would in any 

case reflect the higher levels of 

exposure. 

ATTACHMENT E

PROGRAMMATIC ISSUES TO BE REVIEWED FOR EVALUATING FOR POTENTIAL UNMONITORED INTAKES

Page 13 of 19
Document No. ORAUT-TKBS-0010 5 Revision No. 01 Effective Date: 10/15/2009 Page 139 of 150

Table E-6 Programmatic items potentially affecting the calculation or assignment of dose from plutonium. 

Item Finding/issue description

Dose 

affected DR action Comments

6-1 The Tiger Team Assessment report (DOE 1991) noted that 

commingling of personnel was permitted in break areas in TA-55. 

Pu None No modification of specific dose 

reconstruction assumptions is possible 

based upon this comment. 

6-2 ...eating, drinking, and using tobacco products while wearing protective 

clothing is permitted in some break rooms [in TA-55]. 

Pu None This would increase the amount of 

dose an individual received, but does 

not affect the ability to reconstruct that 

dose. 

6-3 In addition, the method by which workers are chosen for inclusion in 

the bioassay program is inconsistently applied [in TA-55]. (No further 

explanation of ”inconsistent” is given.)

Pu None No modification of specific dose 

reconstruction assumptions is possible 

based upon this comment

6-4 Cracked glovebox gloves contaminated with plutonium were observed 

in TA-21 Building 146 (DOE 1991). 

Pu None Workplace monitoring not used in dose 

reconstruction

Table E-7. Programmatic items potentially affecting the calculation or assignment of dose from plutonium and/or uranium. 

Item Finding/issue description

Dose 

affected

DR 

action Comments

7-1 The Tiger Team Assessment (DOE 1991) indicated that while in some 

areas personnel were either evaluated for participation or all personnel 

participated in appropriate bioassay programs (TA-55 and TA-50), not

all personnel at the plutonium and DU facilities were evaluated for 

participation in bioassay programs. 

Pu and U Evaluate 

claim for 

PUI

7-2 The potential for unmonitored intakes was significant in the early years 

(1944 to 1946) for any site worker. Bioassay was provided only for 

those workers who were most exposed. However, because the 

excretion of plutonium is continuous after an intake, significant intakes 

might have been identified on later routine bioassay samples with 

improved sensitivity for workers who remained at LANL. 

Pu and U Evaluate 

claim for 

PUI

ATTACHMENT E

PROGRAMMATIC ISSUES TO BE REVIEWED FOR EVALUATING FOR POTENTIAL UNMONITORED INTAKES

Page 14 of 19
Document No. ORAUT-TKBS-0010 5 Revision No. 01 Effective Date: 10/15/2009 Page 140 of 150

Item Finding/issue description

Dose 

affected

DR 

action Comments

7-3 As bioassay sensitivities and respiratory protection equipment 

improved, the potential for intakes decreased. Due to rigorous 

workplace monitoring, the probability that a worker received a large 

intake of radioactive material that was unmonitored and unnoticed was 

less after 1946. However, the probability of unmonitored small intakes 

was larger. Periodic reports from H Division of air samples, 

contamination incidents, and hot spots continued to identify a 

significant number of “over tolerance” occurrences throughout the 

1960s. 

Pu and U Evaluate 

claim for 

PUI

7-4 Respiratory protection was available but, except in a few locations, was 

only donned when a CAM alarmed or airborne levels approached 

tolerance or ALs. Thus, a potential existed for an intake before the 

alarm. 

Pu and U None No credit is taken in dose reconstruction 

for respiratory protection.

7-5 Review of Summary of Radiological Incident Reports (RIR) through 

June 1998 indicates that, while the number of RIRs had decreased 

significantly, the potential for unmonitored small intakes continues to 

exist (Bates 1998).

Pu and U Evaluate 

claim for 

PUI

7-6 The Tiger Team assessment (DOE 1991) noted that radiation surveys 

throughout the DU and plutonium areas did not conform to LANL 

policies and documented schedules in TA-55. 

Pu and U None Workplace monitoring not used in dose 

reconstruction

Table E-8. Programmatic items potentially affecting the calculation or assignment of dose from thorium. 

Item Finding/issue description

Dose 

affected

DR 

action Comments

8-1 The Tiger Team Assessment report (DOE 1991) indicated that line 

managers were not aware that thorium and its decay products were 

internal radiological hazards and that workers handling gram quantities 

of dispersible thorium oxide powders and other thorium compounds 

should be identified by the checklist system for participation in the 

bioassay program. 

Thorium Apply 

thorium 

doses 

when 

indicated

Thorium is specifically listed as a potential 

unmonitored actinide source, and 

assignment of a maximum dose is 

described in Section 5.5 of the TBD. 

ATTACHMENT E

PROGRAMMATIC ISSUES TO BE REVIEWED FOR EVALUATING FOR POTENTIAL UNMONITORED INTAKES

Page 15 of 19
Document No. ORAUT-TKBS-0010 5 Revision No. 01 Effective Date: 10/15/2009 Page 141 of 150

Table E-9. Programmatic items potentially affecting the calculation or assignment of dose from plutonium and/or uranium. 

Item Finding/issue description

Dose 

affected

DR 

action Comments

9-1 The 1991 DOE Tiger Team Assessment expressed concern that tritium 

bioassay samples scheduled by the field could remain in the field for 

several weeks rather than being turned in immediately (DOE 1991). 

Tritium NSA Calculated doses would be based upon 

the analysis and the submission dates, 

and positive doses would be backcorrected, but missed doses may have 

resulted when samples dropped below the 

MDA over time. However, this project 

routinely assigns missed dose, so no 

additional action is necessary. 

9-2 Some studies indicate that varying amounts of tritium could be held up 

in the polyethylene container when timely processing of samples (within 

hours) does not occur. 

Tritium NSA This could result in some diffusion of 

tritium through the container, reducing the 

concentration in the sample to some 

extent; however, the potential loss cannot 

be estimated by the dose reconstructor. In 

any case, it is unknown to what extent this 

observation would have affected the 

universe of LANL tritium dose 

measurements. 

9-3 The results are often delayed 2 to 5 weeks after receipt (DOE 1991). Tritium None Delays in communicating results of 

analysis do not affect the dose itself.

9-4 In an evaluation of a job performed while the Tiger Team was 

performing the assessment, an employee at the Ion Beam Facility was 

potentially exposed to substantial quantities of tritium contamination, 

both HTO and OBT. The staff concluded that the facility did not provide 

proper controls and protection or require bioassay analyses for 

potentially exposed personnel.

Tritium NSA Insufficient information is given in the 

finding to project this incident upon a 

population of workers.

9-5 The Tiger Team Assessment also indicated that tritium detectors in 

laboratories were not being source checked (DOE 1991).

Tritium None Dose is calculated without reference to 

workplace monitoring information. 

9-6 The Tiger Team Assessment Report (DOE 1991) indicated that 

calibration and response checking of fixed instruments and tritium 

monitors did not reflect the same level of attention and commitment 

given to portable instrumentation.

Tritium None Fixed instrument results are not used to 

estimate dose from tritium. 

ATTACHMENT E

PROGRAMMATIC ISSUES TO BE REVIEWED FOR EVALUATING FOR POTENTIAL UNMONITORED INTAKES

Page 16 of 19
Document No. ORAUT-TKBS-0010 5 Revision No. 01 Effective Date: 10/15/2009 Page 142 of 150

Item Finding/issue description

Dose 

affected

DR 

action Comments

9-7 Approximately 100 radiation protection instruments were out of 

calibration at one building in the Ion Beam Facility. The reason for the 

instruments not being sent for recalibration was the extended amount of 

time required for the instruments to be returned. 

Tritium None Instrument readings are not used to 

estimate dose from tritium.

9-8 The counting room for tritium smears was in a hallway between two 

rooms with smearable alpha contamination levels in excess of limits of 

DOE Order 5480.11 (DOE 1988).

Tritium NSA Workplace monitoring data, including 

tritium swipe samples, may be used to 

estimate internal doses; however, this is 

not routine, and has not occurred, or has 

occurred with extreme infrequency, on this 

project to date. 

9-9 Vacuum systems and pumps were not contained in enclosures; thus, 

the potential for direct contact with tritium-contaminated oil or airborne 

contamination existed. 

Tritium NSA This project assumes that there is the 

potential for tritium exposure when casespecific information (e.g. tritium dose 

monitoring results) indicate it.

9-10 In addition, LANL did not demonstrate to the Tiger Team staff that 

engineering controls in tritium facilities would minimize internal tritium 

exposure. 

Tritium None Dose control programs do not affect the 

reconstruction of the doses received.

9-11 Tritium air monitors did not have a uniform alarm setpoint and no 

written justifications were available for the alarm setpoints that were 

chosen. 

Tritium NSA Alarm setpoints may affect the dose 

received; however, the finding does not 

specify that the alarm setpoints were not 

compliant with applicable radiation 

protection rules. This project seeks only to 

estimate the dose received with sufficient 

accuracy (to determine probability of 

causation) or, alternatively, to assign a 

maximum dose. 

ATTACHMENT E

PROGRAMMATIC ISSUES TO BE REVIEWED FOR EVALUATING FOR POTENTIAL UNMONITORED INTAKES

Page 17 of 19
Document No. ORAUT-TKBS-0010 5 Revision No. 01 Effective Date: 10/15/2009 Page 143 of 150

Item Finding/issue description

Dose 

affected

DR 

action Comments

9-12 No air samples were collected to quantify the levels of airborne tritium 

contamination in the workplace, except at the Weapons Engineering 

Tritium Facility (WETF), although the potential to exceed 10% of the 

DAC existed. 

Tritium NSA The finding identifies a poor practice that 

may be a violation of applicable radiation 

protection rules; however, this project 

seeks only to estimate the dose received 

with sufficient accuracy (to determine 

probability of causation) or, alternatively, to 

assign a maximum dose.

9-13 Some air monitors were not calibrated at the calibration facility or at all. Tritium 

(comment 

from 

tritium 

facility)

NSA Alarm setpoints may affect the dose 

received; however, this project seeks only 

to estimate the dose received with 

sufficient accuracy (to determine 

probability of causation) or, alternatively, to 

assign a maximum dose.

9-14 The general conclusion of the Tiger Team was that posting and 

contamination control in tritium facilities was nonuniform. This 

nonconservative aspect of the program, coupled with the absence of 

timely tritium smear and bioassay analysis, made the quality of 

radiation worker protection hard to quantify (DOE 1991). 

Tritium None No response to this qualitative comment is 

possible for the purpose of dose 

reconstruction. 

9-15 In addition, the Tiger Team staff was not in agreement with the 

decisions of the management of the Ion Beam Facility about an incident 

that involved a highly contaminated vacuum pump. 

Tritium None No response to this qualitative comment is 

possible for the purpose of dose 

reconstruction.

9-16 The report also expressed concern about the TA-41 treating tritium 

exposures as a lower risk gas rather than tritiated water or organic 

tritium. 

Tritium None Elemental tritium has an ICRP (1978) DAC 

conversion factor 4 orders of magnitude 

lower than HTO, and is thus a lower risk.

ATTACHMENT E

PROGRAMMATIC ISSUES TO BE REVIEWED FOR EVALUATING FOR POTENTIAL UNMONITORED INTAKES

Page 18 of 19
Document No. ORAUT-TKBS-0010 5 Revision No. 01 Effective Date: 10/15/2009 Page 144 of 150

Item Finding/issue description

Dose 

affected

DR 

action Comments

9-17 A concern was expressed that the internal contamination control 

program at LANL relied primarily on administrative controls to maintain 

tritium exposures as low as is reasonably achievable (ALARA) rather 

than obtaining quantitative data from airborne, surface, and bioassay 

sample results.

Tritium NSA Tritium dose monitoring results are present 

in LANL records, and many of them record 

zero doses. From this the inference is that 

the tritium dose monitoring program is 

conservative overall. Environmental tritium 

doses are assigned (or overestimated) for 

all noncompensable claims. 

9-18 Removable and fixed surface contamination limits for tritium and pure 

gamma-emitting nuclides were 100,000 dpm per 100 cm2

. This is not 

in compliance with the limits in DOE Order 5480.11 (DOE 1988).

Tritium 

and 

gamma 

emitters

None Workplace monitoring not used in dose 

reconstruction

ATTACHMENT E

PROGRAMMATIC ISSUES TO BE REVIEWED FOR EVALUATING FOR POTENTIAL UNMONITORED INTAKES

Page 19 of 19
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 145 of 150

Maximizing and Best Estimate Intake Parameters

Intake parameters can be derived from airborne contamination levels for buildings with the highest 

exposure potentials or highest intakes for various periods. Maximum and average airborne 

contamination levels for selected buildings are listed in Table B-1. Incidents and intakes are listed in 

Table F-1. Average airborne contamination levels either are derived as simple averages or are 

reported as averages listed in LANL reports. Simple averaging is assumed for LANL reports, 

although no information on the methods used to obtain these reported averages is available.

Table F-1. Reported exposure incidents and results. Date Incident

1944–1946 26 workers received intakes estimated to be 6 to 80 nCi while involved in plutonium operations: purification, 

fluorination, metal reduction, and recovery. Average intake = 58 nCi (Voelz et al. 1979); 9 of 12 persons with highest 

exposures were working with water-soluble plutonium salts in 1945 (Hempelmann and Langham 1953).

1944 Periodic overexposure of individuals working on RaLa Project occurred for 6 mo at start of project until bugs were 

worked out of remote chemical handling procedures.

1944 Highest nasal swipe for May was 11,372 cpm (assume alpha)

1944 August – Worker opened a sealed container of active material without a respirator or face shield.

1945 August – Building 52 handled large amounts of polonium. Note at bottom of 52 air concentration data table stated:

Note: 0.75 cpm/L is used as 2-year tolerance value assuming 100% retention by lungs."

1945 4 persons exceeded safe amount of plutonium in body, 1 μg (Hempelmann 1946).

1945 Tolerance value for polonium in urine samples 1,500 cpm/d (assume 50% efficiency for counter), exceeded by only 2 

individuals. All persons working with polonium were monitored (Hempelmann 1946).

1945 At the Water Boiler at Omega, several instances of mild to moderate overexposure when gas exhaust lines leaked or 

during decontamination of active material.

1945 Summer – 14 workers removed from plutonium production because urine assays indicated body burdens at or above 

the MPL of 7 cpm/24-hr sample (Hempelmann, Richmond, and Voelz 1973).

1945 Workers entered contaminated side of the air filters areas at DP East without approval (Cox 1945).

1946 Tuballoy fire, Tech Area, April 15; firemen exposed without protective equipment (Tribby 1946b).

1948 November 2 – worker from GMX ingested sufficient amount of RaLa to give a reading on detector (LASL 1948b).

1948 Radioactive specks found near Bayo Canyon ranged from 0.1 to 10 μCi of RaLa at time of origin (LASL 1948a).

1949 Difficulties with RaLa separations and extremely large quantities of material being worked with in Bayo Chemistry 

Area cause spread of contamination and few slight overexposures (LASL 1949b).

1949–1951 Number of airborne contamination results above plutonium MAC (0.0044 cpm/L) decreased from 40% of samples in 

1949 to 7% in 1951 for DP West and 25% to 17% for Tech Area. Above-MPL nasal swipes decreased from 40% in 

1949 to 15% in 1951 for DP West and 60% to 15% for Tech Area (Shipman et al. 1951).

1951 1,876 plutonium urine samples below MAC and 492 polonium urine samples from DP East were below tolerance. All 

routine tritium urine samples were below MAC (Shipman et al. 1951).

1951 DP West dismantle 408 and 413; high airborne plutonium exposures were kept to minimum (Shipman et al. 1951).

1951 DP East experimental program with actinium. No overexposure; however, adequate urine analysis is not available, 

air samples difficult to evaluate (Shipman et al. 1951).

1951 Pajarito Warehouse contaminated by leaking polonium source. TA-33, Area 6 required cleanup from polonium spill 

(Shipman et al. 1951).

1951 Ruptured plutonium slug at GT site vault. Contamination spread to main building. Air tolerance limits reached.

(Shipman et al. 1951)

1951 Airborne alpha activity in Room 513, DP West was consistently above MAC. Source could not be localized; 

however, cleaning attic and sealing openings to room decreased airborne levels to satisfactory (Shipman et al. 

1951). MAC total alpha plutonium as of April 1951 was 0.0044 cpm/L.

1951 Kilocurie quantities of tritium handled in CMR areas. “Health program for this work has been satisfactory” (Shipman 

et al. 1951).

1950s late Radioiodine Experiments: LANL biomedical researchers orally administered I-125 and I-131 to 19 subjects. 

1955 A vial containing americium and curium (91%) exploded, contaminating 2 workers on the face and clothing (LASL 

1955). 

1956 Some high urine counts in uranium areas for employees working with incinerators without respirators. Sigma 

Building, TU Building, Tuballoy Shop, M Building, and Building 96 (McKown 1957).

ATTACHMENT F

DESCRIPTIONS OF SOME INCIDENTS RESULTING IN INTERNAL DOSES

Page 1 of 4 
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 146 of 150

Date Incident

TA-18 air samples in vicinity of Honeycomb assembly were one-half of tolerance level, and in vicinity of Lady 

Godiva assembly air samples were “excessively high.”

May 17 – Spill of Sr-90 within H-7 undetected. Contamination spread to homes (Buckland 1958).

February 12 – Godiva reactor went supercritical and contaminated the kiva (Buckland 1958).

June 13 – Spill of Dy-165 resulted in airborne and personnel contamination (Buckland 1959). Monitoring of 

individual for urine, blood, and human counter showed only positive for the human counter. Suspected the 

human counter to have surface contamination. Worker’s badge read 2 rep (Buckland 1958).

Dismantling tritium equipment in TA-33 resulted in the contamination of three workers, urine concentrations up 

to 650 μCi/L (Buckland 1958)

May 9 – Spill of Dy-165 resulted in airborne and personnel contamination, worker had 20 mR/hr at nostril 

(Buckland 1959).

Total body counts performed on cyclotron personnel indicated measurable Zn-65 (Shipman 1959).

Three employees >MPL, 11 >50% MPL, 90% of people with ≤10% MPL for plutonium (0.04 μCi) of 1,325 

monitored (Shipman 1959)

Radioactive gas leak at LAPRE II resulted in thyroid uptakes by at least five workers of I-131, I-133, and I-135 

resulting in thyroid burdens equivalent to approximately 3 μCi I-131 (Van Dilla 1959).

Laboratory air sample 3.8 dpm/, 16% of MPC for 1-wk sample, highest value February to July (LASL 1959).

Release of plutonium and beryllium due to equipment failure in TA-33 4/13 and 8/4 resulted in intakes and area 

contaminations (Buckland 1961).

March 15 – Wind carried plutonium contamination from an assembly that was being identified to a member of H3. Gross contamination and an intake of plutonium resulted. Contamination spread to the Administration 

building before being detected (Buckland 1961).

March 2 – A power outage disabled exhaust ventilation resulting in the contamination of four laboratories 

(Buckland 1961). 

December 1 – Activated Na-24 in glass wool contamination (Buckland 1961). 

Tritium exposure, 476 μCi/L-urine sample January 11 (Buckland 1961).

Wing 9 of CMR – Inhalation of Pu-238, minimal urine excretion for 100 days, then rose to large values. ICRP 30 

model modified to time constant of 10,000 days (ICRP 1979; Miller et al. 1999); half-time for ICRP 30 model is 

10,000 days, AMAD is 0.2 μm. Intakes 2,150 to 210 nCi, found to fit other Wing 9 intakes.

Demolition of Building 12, DP Mesa, produced no bioassay above detectable and only four detectable nose 

swipes. Highest nose swipe was 85 dpm (Christensen, Garde, and Valentine 1975)

February – One employee received plutonium intake estimated at 1 MPBB (0.04 μCi) (University of California

1977a,d).

January – Oxide of plutonium caught fire, airborne 4.5 to 11.65 dpm/m3, no positive nasal swipes, no bioassay 

(University of California 1977b).

H-3 urine bioassay after January 1 incident at LAMPF <1 μCi/L (LASL 1977a)

February 10 incident resulted in <16-nCi lung burdens for five employees; one employee received DTPA 

chelation therapy. See Table F-2 for details (University of California 1977b). Urine sample plutonium intake 

estimates are 1 nCi (one worker), 4 nCi (two workers), 9 nCi (one worker).

U-235 airborne concentrations of >1,500 dpm/m3 and >25,000 dpm/m3 for DP Bldg 4, Room 412 5/77 

(University of California 1977b).

June – Tritium in urine 101 μCi/L after airborne release DP, Bldg 5.

June – Alpha contamination (up to 100,000 cpm) from damaged uranium target found on several persons at 

LAMPF switchyard (University of California 1977d).

May – TA-35 and TA-41 alpha contamination escaped from pressure vessel, caused personnel contamination 

(University of California 1977d).

20,000 dpm alpha airborne (U-235 and U-238 operations). 

November – Tritium bioassay (confirmed) of P-9 employee 798 μCi/L (University of California 1977c).

UF6 tube ruptured (5-10 g) spreading HF fumes and depleted U-238 (University of California 1978a).

Operator replacing leaking hose at TA-33 stack had tritium bioassay of 4 μCi/L; 449 Ci HT released (University 

of California 1978b).

Zia employee with nasal swipe of 553/70 dpm TA-21 (LASL 1980a).

TA-33 tritium exposure to employees.

54-μCi/L tritium bioassay TA-33 (LASL 1980b).

153-μCi/L tritium bioassay TA-33 (LASL 1980c).

Plutonium spill in TA-55, 10 persons, no significant exposures (LANL 1983a).

ATTACHMENT F

DESCRIPTIONS OF SOME INCIDENTS RESULTING IN INTERNAL DOSES

Page 2 of 4 
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 147 of 150

Date Incident

High airborne contamination results in high nasal swipes TA-55 plutonium (LANL 1983b).

21 plutonium wound counts, 3 = 0.1 nCi, 1 had 11.2 nCi Pu-239 and 0.4 nCi Am-241 and was surgically excised 

(LANL 1983c).

Quarter 4 – One wound above 20 nCi, numerous positive (>50 dpm) nasal wipes (LANL 1983b)

Particulate filtering respirators were available and were used from the beginning of the program. 

Therefore, ambient air concentrations might not reflect the actual breathing air concentration of the 

workers.

Research continues to identify the intakes for both maximizing and general conditions. Examples of 

intakes identified thus far are listed in Table F-1; Table F-2 summarizes a particular incident. As the 

state of the respiratory protection program and engineering controls improved, the potential for 

chronic and acute intakes was reduced.

Table F-2. Summary of February 10, 1977, plutonium incident (University of California 1977d).

Worker

Total systemic burden (increase) Nasal swipe (L/R) dpm Chest burdena

Pu-239 nCi Pu-238 nCi Total alpha Pu nCi

Technician 1 8.2 (3.3) 1.7 (0.1) 12,759/13,385b 7.5

Technician 2 7.6 (5.8) 0.7 (0.5) 801/1,849 13.

Supervisor 2 42.9 (7.5) 3.2 (0.6) 1,091/1,193 6.

Technician 3 21.8 (18.0) 1.1 (0.4) 611/1,193 10.

HP Surveyor 4.2 (0.3) 0.7 (0.0) 226/577 0

a. No worker had chest counts above MDA (16 nCi).

b. Received DTPA chelation therapy.

NOTE: While an attempt has been made to report only incidents with quantitative results, some 

incidents for which only qualitative comments were available in the records have been included in the 

list. Information is not available for every year of operation.

Table B-2 lists maximum airborne concentrations for selected buildings during various intervals. In 

practice, individuals in areas of maximum airborne concentration would probably have worn 

appropriate respiratory protection.

Rather than preventing intakes completely, the focus of the respiratory protection program, especially 

in the early years of operation, was to not exceed allowable concentration limits. The MPBB for 

plutonium was 5 μg from 1943 to 1945. During 1945, when many of the exposures occurred, the 

tolerance dose of plutonium was assumed to be 1 μg. Therefore, a suitable method of measuring 

personnel exposure had to be able to detect from 2 to 14 dpm of plutonium in a 24-hour sample of 

urine.

The MACs for buildings with a high potential for exposure were established by 1947 and changed 

over the years as technology and understanding of the hazards increased. The alpha MAC values 

used for selected buildings are listed in Table B-2. Assume that the period for the listed MAC 

continues until changed. The MACs are dpm for total gross alpha with the primary nuclide for the 

area listed, except as noted. MACs were eventually replaced by maximum permissible 

concentrations (MPCs) and derived air concentrations (DACs) as controlling values. Los Alamos ALs 

are also listed if applicable. It appears that at some time before 1964, the MAC values were changed 

ATTACHMENT F

DESCRIPTIONS OF SOME INCIDENTS RESULTING IN INTERNAL DOSES

Page 3 of 4 
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 148 of 150

to ALs with larger, more specific MAC values established for nuclides, enrichments, and mixtures. 

Table B-3 includes an example of the ALs and MACs for selected buildings in 1964.

ATTACHMENT F

DESCRIPTIONS OF SOME INCIDENTS RESULTING IN INTERNAL DOSES

Page 4 of 4 
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 149 of 150

Scaling of Assigned Intakes Using Case-Specific Information

The major presumptive exposures at LANL were made up of the primary radionuclides plutonium, 

uranium, tritium and, in early days, polonium. Other radionuclides may have resulted in intakes to 

personnel, but due to the small amounts of material, the limited operations, and the use of 

engineering controls such as dry boxes, potential exposures are relatively unlikely. When exposures 

are possible, the intakes were likely to be small, and highly dependent upon work category. Work 

locations are another important consideration when evaluating for a potential unmonitored intake. 

Though it is not possible in all cases to precisely locate LANL employees, it is certainly true that the 

non-primary radionuclides presented a limited hazard to most LANL personnel. In this Attachment, a 

method is presented to scale assigned intakes based on job category and likely duration of exposure 

in order to limit the margin of overestimation.

Level of contact. For the purpose of dose reconstruction, the amount of exposure should be 

considered when assigning coworker dose intakes. For personnel with full contact with the material, 

the 95th percentile of the calculated coworker intakes should be assigned. For others, the 50th

percentile is more appropriate, as listed in the table below. 

Table G-1. Levels of contact for certain job categories. 

Contact 

level Description Example job titles

Percentile of 

CW Intake

Primary 

contact

Contact with material in unencapsulated 

and uncontained form. Normal contact 

with the material is under less-controlled 

environments such as in fume hoods 

and inside chemical apparatus.

Chemists, researchers 95th

Secondary 

contact

Normal contact with the material is in 

containment or process equipment, 

occasional contact with airborne 

material during excursions. 

Operators, janitors, 

decommissioning and 

decontamination workers, health 

physics personnel routinely 

assigned

50th 

Incidental 

contact

Contact with contamination and airborne 

radioactive material only during an 

excursion or incident or as part of an 

exposure during a maintenance, repair, 

waste disposal, decontamination, or 

decommissioning activity.

Janitors decommissioning and 

decontamination maintenance, 

health physics personnel not 

routinely assigned construction 

activities associated with waste 

disposal or demolition of legacy 

structures

50th

Contact 

unlikely, but 

presumed

Nonradiological job categories,

construction activities not associated

with waste or decontamination and 

decommissioning, but evidence 

supports a potential incidental exposure.

Construction activities 50th

Contact 

unlikely, not 

presumed

Nonradiological job categories, and no 

evidence to suggest a presumptive 

exposure.

50th

ATTACHMENT G

SCALING OF COWORKER DOSE INTAKES CONSISTENT WITH 

TYPE AND DURATION OF EXPOSURES

Page 1 of 2
Document No. ORAUT-TKBS-0010-5 Revision No. 01 Effective Date: 10/15/2009 Page 150 of 150

Dose Assignment. Fiftieth-percentile intakes are used to calculate organ doses based on the listed 

lognormal distribution of intakes, along with the listed geometric standard deviation. Ninety-fifthpercentile intakes are assumed to be constant values (distribution-free). The actual intake values are 

calculated using the relationship below. 

95th-percentile intake = 50th-percentile intake X GSD1.645

ATTACHMENT G

SCALING OF COWORKER DOSE INTAKES CONSISTENT WITH 

TYPE AND DURATION OF EXPOSURES

Page 2 of 2